ML100740573
| ML100740573 | |
| Person / Time | |
|---|---|
| Site: | University of Wisconsin |
| Issue date: | 10/17/2008 |
| From: | Agasie R Univ of Wisconsin - Madison |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| COWDREY, CHRIS, NRR/DPR/PRTA 415-2758 | |
| References | |
| TAC ME1585 | |
| Download: ML100740573 (5) | |
Text
{{#Wiki_filter:UNIVERSITY OF WISCONSIN NUCLEAR REACTOR LICENSE NO. R-74 DOCKET NO. 50-156 LICENSE RENEWAL APPLICATION SAFETY ANALYSIS REPORT, AND TECHNICAL SPECIFICATIONS REVISION 2 REDACTED VERSION* SECURITY-RELATED INFORMATION REMOVED
- REDACTED TEXT AND FIGURES BLACKED OUR OR DENOTED BY BRACKETS
Nuclear Nuclear Reactor Reactor Laboratory Laboratory UWNR University of Wisconsin-Madison UWNR Wisconsin-Madison 1513 1513 University Avenue, Room 1215 1215 ME, Madison, 53706-1687, Tel: (608) Wl 53706-1687, Madison, WI (608) 262-3392, 262-3392, FAX: (608) (608) 262-8590 262-8590 reactor@engr.wisc.edu, http://reactor.engr.wisc.edu email: reactor@engr.wisc.edu, http://reactor.engr.wisc.edu October 17, 17, 2008 981 RSC 981
- Director, Director, Nuclear Reactor Reactor Regulations Regulati~ns ATTN:
ATTN: Document Control Control DeskDesk u.s. U.S. Nuclear Nuclear Regulatory Regulatory Commission Commission Washington, DC 20555 Washington, 20555
Subject:
Subject:
License License Renewal Renewal of the UniversityUniversity of Wisconsin Nuclear Nuclear Reactor; Reactor; License R-74, R-74, Docket 50-156
Dear Sir:
The attached information information is submitted in is submitted in support of our application for license renewalrenewal dated May 9, 9, 2000, 2000, as supplemented supplemented on May 6, 6, 2004 and September September 2, 2, 2004. 2004. As a result.result of the extensive:extensive changes, changes, a complete replacement replacement copy of the Safety Safety Analysis Report for Renewal of of License License R-74 for the University of Wisconsin Nuclear Nuclear Reactor Reactor is is enclosed. enclosed. The changes described in changes described in the attachment are indicated indicated by a vertical vertical line i~in the margin of the enclosed enclosed revision revision 2 to the Safety Analysis Analysis Report Report for RenewalRenewal of License License R-74 for the University University of Wisconsin Wisconsin Nuclear Nuclear Reactor, September 2008. Reactor, dated September 2008. In accordance In accordance with 10 CFR 50.30(b), 50.30(b), I declaredeclare under penalty penalty of perjury that the foregoing is is true true and correct. correct. If you have any further questions, If questions, please,contact please contact me at (608) (608) 262-262-3392. 3392. Robert J. ggas/e Director Reactor Director Attachment Attachment Enclosure Enclosure AC)D Z) cc: Daniel Hughes,Hughes, NRC Project Manager Manager
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SAR Revision 2 changes Page 1I of 14 Section Rev2 Page Change Change Reference Reference Chapter 11 Chapter 1.2 1-1. 1-1 Fixed typo. 1.3 1-2 1-2 UpdaJ~d excess reactivity Updated reactivity (4.9 to 4.3) 4.3) and shutdown shutdown margin (4 to 4.2) to reflect current core measurements. 1.3 1-2 Changed "TRIGA-FLIP Changed "TRIGA-FLIP High" High" to "TRIGA-FLIP,"
"TRIGA-FLIP," fixing typo.
1.3 1-3 Fixed typos. 1.3 1-6 Expanded Expanded "STD fuel" "TRIGA Standard fuel" fuel" to "TRIGA fuel" for clarification. 1.4 1-6 Eliminated mention of sub-critical sub-critical assembly and hotcell, since the hotcell is now - J.4 contained within the restricted area and the sub contained subcritical critical assembly assembly has been removed. 1.4 1.4 1-6 Corrected Corrected facilities wording since now all HVAC HV AC is non-shared use except offices. 1.6 1-7 Removed sentence about iodineiodirie and strontium-90 strontium-90 inventory inventory limits; iodine limits are covered in Technical Specification covered Specification 14.3.8.2, 14.3.8.2, and limits on all fission products products are covered in section 11.1.1.2, 11.1.1.2, page 11-3, 11-3, which states states that all activity must be below 10 CFR 20 limits for effluent effluent when assuming assuming a months dilution with the ventilation ventilation
- system.
tn tn 1.8 1-8 Corrected initial criticality date from January 55 th Corrected 2 6th 1961. to March 26 1961. It is not clear th if anything significant significant occurred on January 5th, , but it was not the initial criticality criticality date as reported in chapter 4 of the SAR and as printed on the dedication dedication plague displayed in the control room. 1.8-1.8- 1-8 Clarified Clarified that the initial replacement replacement of 9 Standard Standard fuel bundles with FLIP was for a mixed core. 1.8 1-8 Fixed typo by adding "was critical in March 1974".
"was initially critical 1974".
1.8 1-8 Updated current MW Hours Hours to over over 20,000 (20,030 as of 9/16/2008) 9/16/2008) and added units
--
of MWd. ofMWd. 1.8 1-8 Updated Updated most recent amendment amendment number to 16 dated August 30,2006. 30, 2006. 1.8 1-8,9 modification history. Updated facility modification ---
SAR Revision 2 changes Page of 14 Page 22 of 14 Section Rev2 Page Change Reference Chapter 22* 2.1.1.1 2-1 Updated Madison population population with with 2000 census data (changed from 191,000 to 208,000). 208,000). 2.1.1.1 2.1.1.1 2-1 Added clarification Added clarification that 130m distance to nearest residence is 80m from building wall to avoid confusion with LEU report analysis for ground ground release. 2.1.1.2 2.1.1.2 2-1 Changed reference Changed reference from Figure 2-4 to 2-5, 2-5, typo. 2.1.1.2 2.1.1.2 2-1 Updated reactor lab room number number from 130 to 1215, 1215, and clarified the site boundary _definition. definition. tn tn 2.1.1.2 2.1.1.2 2-1 Added reference to ME floor maps for 4 4 th and 5th floors. Also added reference reference to new building cross section figures. 2.1.2 2-2 Updated to 2000 census data. Newreference New reference has replaced replaced original reference reference 1. Response to NRC
Response
comment #4 ()(I) comment 2.1.2 2-2 Updated ME building occupancy occupancy estimate. The initial estimate was 1182 assuming all lecture halls, computer labs, and offices offices were fully occupied, but this does not account for laboratories account laboratories (such as engine engine labs), or hallways/lobbies hallways/lobbies (such (such as with Engineering Engineering Expo). Therefore the initial estimate Expo ). Therefore estimate was rounded up to 1500. The previous value was 300. 2.1.2 2.1.2 2-2 2-2 . Updated current current Camp Randal capacity capacity from 76,129 to 80,321. 80,321. Source: UW Badger Badger Athletics website (http://www.uwbadgers.com/facilities/camprandall/index_38.html). (http://www.uwbadgers.comlfacilities/camp randall/index 38.html). Figure Figure 2-3 2-3 Updated Madison Madison city map using Microsoft circles had been Microsoft Virtual Earth. Distance circleshad been requested by NRC. Figure 2-4 2-4 Updated campus map using www.map.wisc.edu. www.map.wisc.edu. Figure 2-5 2-5 Updated engineering engineering campus map using www.map.wisc.edu. www.map.wisc.edu. Figure 2-6 2-6 2-6 Updated Updated ME basement basement map. Figure 2-7 2-7 2-7 Updated ME first floor map. Figure 2-8 2-8 Updated ME second floor map. Figure 2-9 2-9 Updated ME third floor map. Figure 2-10 2-10 Added new ME fourth floor map.
SAR Revision 2 changes changes . Page 3 of 14 of 14 Section Rev2 Rev2 Page Change Change Reference Reference Figure 2-11 2-11 Added new ME fifth floor map. Figure 2-12 2:..12 2-12 Added new ME cross-section cross-section map looking north. Figure 2-13 2-13 2-13 Added new ME cross-section map looking east. Table 2-1 2-14 Updated with 2000 census data. Corresponding Corresponding reference reference 1 has also been updated. Response Response to NRC comment #4 (1) () 2.2.1 2-14 Updated National Guard annual flight missions from 1999 figure of approximately approximately 3000 to 2006 figure of approximately 3000to approximately 4000 (exact number is 8130 events). 2.2.2 2-15 1999 traffic statistics with 2006 numbers for total events and Updated 1999 commercial/general/military corrected typo with runway commercial/general/military percentages. Also corrected headings (3100 (310° should have been been 3200). 320°). Figure 2-12 2-22 Updated Madison Well figure and made it full-page to make it easier to see. 2.6 2.6 2-26 reference 1. Updated reference 1. Response Response to NRC comment #4(1) comment #4() 2.6 2-26 references 33 and 4 (national guard and airport telephone Replaced references telephone interviews) interviews) with reference 3 (email interview with airport). All later references new reference references were renumbered and the corresponding footnote labeling in the body of the chapterchapter was also renumbered, renumbered, but these renumbering renumbering changes changes were not redlined.
-
Chapter 33 3.1 3-1 Added history of ME construction, construction, new ventilation system and cooling system, design criteria. 3.2 3-1 Updated 40 year history Updated40 history to 50. Cha~ter4 Chapter 4 4.1.1 4-1 Updated reactivity (4.2 to 4.3) and shutdown margin (4 to 4.2) to reflect Updated excess reactivjty current core measurements. measurements. 4.1.2 4-3 Fixed typos. 4.1.2 4.l.2 4-5 4:'5 Updated excess reactivity (4 to 4.3) and reactivity in control control blades 7.1) to blades (6.9 to 7.1) reflect current core measurements. core measurements.
SAR Revision SAR Revision 22 changes changes Page Page 4 of 14 Section Section Rev2 Page Rev2 Pa~e Chan~e Change Reference Reference 4.1.3.# 4.1.3.# 4-5,6 4-5,6 Added outline Added outline numbering. numbering. 4.2 4.2 4-7 4-7 Added reference Added reference to to 10 10 CFR CFR 73.2 for for-definition quantity HEU. definition of formula quantity HEU. 4.2.1 4.2.1 4-8 4-8 Removed reference Removed reference to aluminum aluminum irradiation irradiation thimble in in 3-element 3-element fuel fuel bundle bundle assemblies, as assemblies, as they are nono longer there any plan to use them in used nor is there longer used in the future. future. 4.2.1 4-8 4-8 Corrected Figure Corrected Figure 4-2 reference, reference, fixing typo. typo. Response to NRC
Response
, #7 (1) comment #7 (\) ,
4.2.2.# 4.2.2.# 4-13,14,15 4-13,14,15 numbering. Added outline numbering. 4.2.2.1 4-13 4-13 Corrected Figure Corrected Figure 4-2 reference, fixing typo. Response to NRC
Response
comment #7 (1) comment #7 (I) 4.2.2.2 4-14 4-14 Clarified wording Clarified wording of safety safety blade blade section. 4.2.5 4-20 4-20 Fixed typo. 4.2.5 4-23 4-23 Added reference reference to Table 4-1 and 4-2 to explain codes in Figure Figure 4-15. 4-15. 4.3 4-26 4-26 history to 50. Updated 40 year history Updated 50. 4.4 4-26 4-26 Included reference Included reference to section calculation of dose rates and integrated section 13.1.3.2 for calculation integrated rd dose for 33 rd floor occupant. Stated that with evacuation, evacuation, public dose would be less less (' than 100 mrem (about 13 mrem). t 4.4.# 4-27 Added outline numbering. 4.5.1 4-27 Sections 4.5.1 and 4.5.2 were were already combined, combined, but outline numbering was revised to indicate section 4.5.1 rather than 4.5.112 4.5.1/2 for clarity. 4.5.1.# 4-28,29,32, 4-28,29,32, Added outline numbering.
,
34,37,40, 44,47,48 4.5.1.2.1 4-32 Fixed typo (reference to Figure 4-15) 4.5.1.3.1 4-37 Fixed typo. 4.5.1.3.2 4-43 Fixed typo. 4.5.1.3.3 4-44,46,47 Fixed typos. 4.5.2 4-51 Updated numbering from 4.5.3 to 4.5.2 (due to combination of previous sections) I _____________ i
SAR SAR Revision 2 changesPae5ol changes Page 55 of 14 Section Section Rev2 Page Rev2 Page Change Change Reference Reference Chapter Chapter 5 5 5.5 5.5 5-3,4 5-3,4 Split sections 5.6 (water cleanup sections 5.5 and 5.6 cleanup and makeup sections) an4 makeup sections) for clarity. Also added system to section 5.5. added description of waste system Previous Figure 5-2 5.5. Previous 5-2 has also been been updated and split into Figures 5-2, 5-3, Figures 5-2, 5-3, 5-4, and 5-5 5-5 to separate separate pooi, pool, cleanup, cleanup, makeup, and waste systems. 5.5 5.5 5-3,4 Updated flow rate and valve numbers for new demineralizer Updated demineralizer (18(18 gpm, valve 10102 10 102 service service out, valve valve 22). 5-2 Figure 5-2 5-5 5-5 Added new pool water schematic. water systems schematic. Figure 5-3 5-3 5-6 5-6 Added new water cleanup system Added system schematic. schematic. Figure 5-4 5-7 5-7 schematic. Added new water waste system schematic. Figure 5-5 5-5 5-8 5-8 Added new water makeup Added schematic. makeup system schematic. ' . Chapter 6. Chapter 6. 6.2.1 6.2.1 6-1 6-1 Updated reference Updated account for added Is' reference to figures to account 151 floor plan. plan. 6.2.1 6.2.1 6-1 6-1 Changed Changed word roof to ceiling to clarify that the confinement confinement ceiling ceiling is no longer the building roof. 6.2.1 6-1 6-1 Eliminated of the console Eliminated mention ofthe conditioner and updated description console air conditioner description to reflect reflect new ventilation system. 6.2.1 6.2.1 6-1,2 6-1,2 Updated Updated lab description to reflect construction construction changes, changes, elimination elimination of windows, change of doors, etc. Security change Security details of doors and window are not specified specified due to
-SGI. SOL Changed Changed name of basement basement area area *from "Nuclear Engineering
.from "Nuclear Engineering Laboratory" Laboratory" to "Reactor Laboratory "Reactor auxiliary support space."
Laboratory auxiliary space." Deleted Deleted paragraph paragraph about about future plans for room on north of lab. lab. 6.2.2 6.2.2 6-2 6-'2 Corrected typo in outline Corrected numbering; 6.2.3 outline numbering; 6.2.3 should have 6.2.2. have been 6.2.2. Figure 6-1 Figure 6-1 6-3 6-3 Updated lab basement Updated basement floor plan. Figure 6-2 Figure 6-2 6-4 Added new lab first floor plan. Figure 6-3 Figure 6-3 6-5 6-5 Updated diagram of lab facing south. Updated ________ Figure Figure 6-4 6-6 6-6 Updated diagram Updated diagram of lab facing facing north.________ north. Figure 6-5 Figure 6-5 6-7 6-7 Updated diagram of lab facing east.' east. ________
SAR Revision 2 changes changes Page 66 of Page of 14 Section Section Rev2 Page Change Change Reference Reference Figure 6-6 Figure 6-6 6-8 Updated diagram Updated diagram of of lab lab facing facing west. Chapter 7 Chapter 7.1 7-1 Updated to include Updated include digital fuel temp, computer pulse, all temp, computer all digital recorder. Figure 7-1 Figure 7-1 7-2 Updated Updated figure to reflect reflect console upgrade. RSC 773 (2 (L) 803 (3) RSC 803 (3) RSC'887 RSC 887 (4) (4) 7.2.3.# 7.2.3.# 7-3,4 Added outline Added outline numbering. numbering. 7.2.3 7.2.3 7-3 Corrected references to bistables Corrected references which should bistables which should be relays. 7.2.3.3 7.2.3.3 7-3 Clarified pulse output output on console computer to reflect new pulse channel. console computer channel. 7.2.3.5 7.2.3.5 7-4 Clarified console recorder. 7.2.4 7-4 Updated history to 50 years. Updated RSC 887 (4) 7.2.5 7.2.5 7-5 Removed reference reference to LogN not to LogN in operate not in operate scram which no scram which longer exists. no longer exists. RSC 887 (4) 7.3.# 7.3.# 7-5,6,8,9 7-5,6,8,9 outline numbering. Added outline 7.3.3 7-5 Corrected Corrected typo, "PULSE" "SQUARE WAVE" "PULSE" position to "SQUARE WAVE" position. 7.3.4 7-6 Clarified pulse output reflecf new pulse channel. computer to reflecfnew output on console computer channel. which was scram which operate scram removed. RSC 887 (41 7.4 7-10 Removed reference reference toto LogN not in LogN not in operate was removed. RSC 887 (4) 7.4 . 7-10 level alarm is on high or low. Clarified that pool level upgrade. RSC RSC 887 (4) 887 Figure 7-2 Figure 7-11 Updated per console console upgrade. (4) upgrade. RSC RSC 887 (4, 887 (4) Figure 7-3 Figure 7-12 Updated per console console upgrade. 7.6.# 7-13,14,15 Added outline outline numbering. 7.6.1 7-13 Added note that area radiation level high alarms at UWPD and initiates evacuation. evacuation. 7.6.1 7-13 Added Evacuation Evacuation Alarm Alarm in Local annunciator. RSC 893 (:l) (5) 7.6.1 7-13 Updated SAM/CAM SAM/CAM troubletrouble annunciator. RSC 896 (6) (b) 7.6.1 7-13 Added Loss of Off-Site Off-Site Power annunciator. RSC 856 (I) (7) 7.6.1 7-14 Removed pn blower annunciator. RSC (8) 857 (IS) 7.6.1 7-14 7-14 Updated to reflect UWPD name change. 7.6.3 7-14 Updated pneumatic system panel description. RSC 857 (IS)7 7.6.4 7-15 Updated ventilation ventilation system panel description, description, including new BP&TC EF-17. (19 RSC 879 ('1)
SAR Revision 2 changes SAR changes Page 77 of 14 Section Rev2 Page Change Reference Reference 7.6.5 7-15 Added reference reference to chapter chapter 5 for details details of cooling system. 7.6.6 7-15 Added whale system panel description description for consistency. 7.7.# 7.7.# 7-15,16 7-15,16 Added outline outline numbering. 7.7.2 7-16 Updated stack stack air monitor description. air monitor description. RSC 896 RSC 896 16, to)
~
Chapter 8 All This entire chapter chapter has been rewritten to reflect changes due to ME building construction, going into greater detail than previously. Also added electrical drawings. 8.2 8-3 Added paragraphs describing UPS per facility mod, but also expanded Added paragraphs expanded the last RSC 895 (IU)0 sentence to reiterate that the UPS is not required "for "for maintaining-the maintaining the facility in safe safe shutdown, even for extended extended periods of time." time." Typo in RSC RSC document referring referring to 6000 kVA kVA has been corrected corrected to 6000 V A. VA. Chapter 99 9.1 9-1,2,3,4 Rewrote Rewrote ventilation system description. This is a complete rewrite which is based t~) RSC 879 (9) on the RSC description. 9.1L# 9.1.# 9-1,2,3,4 Added outline outline numbering. Figure 9-1 9-5 Updated Updated for new vent system. t~) RSC 879 (9) Figure 9-2 9-6 Added Added new figure 9-2 for cross-section ventilation ventilation layout. Figure 9-3 9-7 Added Added new figure 9-3 for cross-section ventilation ventilation layout. Figure 9-4 9-8 Revised Revised fuel rack diagram to reflect actual locations, and added North legend. 9.2.2 9-9 Deleted potentially potentially sensitive information on un-irradiated un-irradiated fuel storage. 9.2.3 9-10 Corrected Corrected typo. 9.2.3 9-10 Added Added new dummy element with reduction AND increase increase in diameter. RSC 869 '*(II)
*
(12) RSC 919 (12) 9.3 9-13 Updated fire systems~ Updated systems. RSC 894 t1j) (') 9.3 9-13 Updated P&S to UWPD UWPD notation. 9.4 9-13 Changed Changed wording wording to auxiliary support space for consistency. na 9.4 9-13 Removed description of 22nd Removed which was uninstalled. intercom system which
SAR Revision 2 changes changes Page 88 of 14 Section Rev2 Page Change Reference Reference 9.5 9-13 Removed old rooms, updated updated lab room number to 1215, 1215, basement basement area room
. number to B1215, B 1135.
B 1215, and added BI135. 9-14 Updated Agreement Agreement State license license number. Chapter 10 10 10.2.# 10.2.# 10-1,3,6,8 10-1,3,6,8 Added outline outline numbering. 10.2.2 10-3 "beams" to "beams Revised "beams" "beams of radiation" for clarification. clarification. Figure 10-2 10-3 Updated Figure 10-2 to include core box and actually label TC and beam ports. 10.2.3 10-6,7 Updated for new pneumatic pneumatic system and modified modified to reflect reflect the basement basement move. RSC RSC 857 (K) 857 (8) 10.2.3 10-6 Clarified wording of activity limits. of activity limits. RSC RSC 879 879 (9) ('.I) Figure 10-4 10-7 Updated for new pneumatic system. RSC 857 (K) ( 10.2.4 10-8 Deleted description of old smaller smaller hydraulic hydraulic irradiation facility. 10.3 10-11 Changed "if released released with 30 days of dilution" to "when averaged over over 30 days of of dilution".. dilution" 10.3 10-11 10-11 Removed outdated outdated reference reference to 130 vs 131 activity activity limits (they are now the same). (9) RSC 879 ~'I) 10-11 10-11 Changed 1 SROSROto to 22 SROs to reflect new commitment to SROs to reflect new commitment to RSC. RSC. Charter RSC Charter (14) RSC 10.3 (14) 10.3 10-11 10-11 Fixed gender language. Chapter 11 11 11 11-1 Updated Agreement Agreement State license number. 11 11-1 11-1 Updated reference reference to radiation safety regulations regulations to remove remove urluri (since this has changed and may change again). 11.1.1.1 .I 11.1.1.1.# 1 1-1,2 11-1,2 Added outline outline numbering. 11.1.1.1.2 11.1.1.1.2 11-2 Updated Ar-41 release release calculations calculations for new vent system. (9) RSC 879 ('.I) 11.1.1.1.2 11.1.1.1.2 11-3 Deleted redundant redundant sentence sentence that volatile or powder powder would be of importance. 11.1.1.1.2 11.1.1.1.2 11-3,4 Updated dilution assumptions assumptions for new vent system. (9) RSC 879 ('.I) 11.1.1.1.2 11.1.1.1.2 11-3 11-3 Updated descriptions descriptions for single fume hood. 11.1.1.1.2 11-4 Clarified wording on RSC approval of non-routine non-routine samples. RSC .879 (9) 879 ('.I) 11.1.2/3 11-5 Added note that these 2 sections are combined.
SAR Revision 2 changes changes Page 99 of 14 Section Section* Rev2 Page Page Change Reference Reference 11.1.5 11-5 Changed students to experimenters to students experimenters reflect reflect that we will no longer TLD badge all RSC 922 ((D) students. 11.1.5 11.1.5 11-5,6 11-5,6 Paragraph about high radiation levels in experiments was revised to reference access Paragraph access control and postings accordance with 10 CFR 20.1601. po stings in accordance 20.1601. 11.1.5 I 11-6 Tour/visitor dose paragraph paragraph was revised to reflect current current use of non-radiation non-radiation worker worker classification classification (based on University Radiation Radiation Safety Regulations). Regulations). 11.2.2 11.2.2 11-7 Added radioactive sink to list of liquid wastes to holdup tank. 11.2.2 11.2.2 11-7 parenthetical reference Included ,parenthetical reference to short description of waste holdup tank in section 5.5. 11.2.3 11.2.3 11-7 The filter size for liquid waste discharge discharge was supposed to be 0.5 micron (0.4 micron was a typo). The UWNR has always used 0.5 micron micron filters, as approved by the Reactor Reactor Safety Committee. 10 CFR 20.2003 says that waste waste must be readily soluble. NUREG-1556 NUREG-1556 Vol. 7provides clarification of "readily 7 provides clarification soluble" by "readily soluble" referencing referencing Information Notice 94-07, but there is no guidance guidance on what size filters to use to remove any potential dissolved solids. UW Radiation Radiation Safety was consulted and stated that 0.5 microns microns is the common filter size used to remove any possible possible - dissolved solids in order to comply with 10 CFR 20.2003.20.2003. 11.3 11-7 Removed website location from reference reference 1, since this has changed changed and may change . again in the future. Chapter 12 Chapter 12.1.2.# 12.1.2.# 12-1,2,3 12-1,2,3 Added outline numbering. 12.1.2.2 12.1.2.2 12-1 Updated Updated department name that Radiation Safety is part of (now University Department Department of Environment, Health and Safety). 12.1.2.3 12.1.2.3 12-1 Department Chair. ANS 15.4 provides very little detail for Added responsibilities of Department any of the levels. Level Levell1 just says "Individual "Individual responsible for the reactor reactor facility's facility's licenses or charter." charter." In our case the chair is also responsible for appointing the reactor reactor director. 12.1.2.6 12.1.2.6 12-3 Fixed gender language.
SAR Revision SAR Revision 22 changes changes Page 1010 of 14 Section Section Rev2 Page Rev2 Page Chan~e Change Reference Reference 12.1.3 12.1.3 12-4 12-4 Corrected outline Corrected outline numbering numbering typo. 12.1.4 12.1.4 12-4,5 12-4,5 description to reflect Revised description reflect current current Operator Candidacy Program. Also spelled Operator Candidacy OJT as On out OJT On the Job Job Training Training for clarification, clarification, and and removed course urI in case course url of case of future change. change. 12.1.5 12.1.5 12-5 12-5 Updated department Updated department name name that that Radiation Radiation Safety Safety is part of (now University University Department of Environment, Health Department Health and and Safety). 12.2.1 12.2.1 12-7 12-7 Radiation Safety name. Updated Radiation 12.2.3 12.2.3 12-7 12-7 gender language. Fixed gender language. 12.2.4 12.2.4 12-8 12-8 Radiation Safety Updated Radiation Safety name. 12.3 12-9 12-9 Changed 1 SRO Changed SRO to 2 SROs required required for temporary temporary procedure fixed procedure changes, and fixed gender language. gender language. 12.4 12-10 12-10 gender language. Fixed gender language. Chapter 13 Chapter 13 13.1 13.1 13-. . 13- Sections 13.1 and 13.2 Sections 13.2 were were previously previously combined, combined, but the outline numbering numbering has 1,2,3,6,8-1,2,3,6,8- 13.112 notation. Also added note to clarify now been revised to eliminate 13.1/2 12,14-16 12,14-16 combination of sections. combination 13.1.1.# 13.1.1 .# 13-2,3,6 13-2,3,6 Added outline numbering. Added 13.1.1.4 13-3 13-3 Added clarification clarification on assumed evacuation evacuation time (5(5 minutes to exit confinement, confinement, another 5 minutes minutes to exit building). Also added reference to Appendix A for clarification, and added sub-section sub-section numbering. 13.1.1.4 13-3 Revised self-contained self-contained breathing apparatus apparatus to powered powered air purifying purifying respirator. Table 13.1 13-4 13-4 Updated H and J values for new vent system. RSC 879 l(9)
'l) 13.1.1.5 13.1.1.5 13-6,7 13-6,7 Updated for new vent system and made minor clarifications. l RSC 879 (9)
'l) 13.1.2.#
13.1.2.# 13-9,10 13-9,10 Added outline numbering. 13.1.3.# 13.1.3.# 13-13-11,12,14 Added outline numbering. . 11,12,14
SAR Revision 2 changes I11 of 14 Page 11 Section Rev2 Page Change Reference 13.1.3.1 13-11 13-11 Updated pool drain time calculations. There There is no record of how the previous calculation was performed, but current current calculation calculation uses reasonable reasonable methods and assumptions with references. references. ro 13.1.3.2/3 13.1.3.2/3 13-12-14 13-12-14 unshielded core dose calculations. The calculated 33rP Updated unshielded floor classroom classroom dose-rates were significant, so an additional additional analysis analysis was performed to model the integrated dose received received to ensure it would be less than 1100mrem00mrem during the evacuation. Integrated dose was calculated calculated to be about 13 mrem. This sectionsection was also divided into. 13.1.3.2 13.1.3.2 for confinement confinement doses, and 13.1.3.3 for unrestricted area area doses. 13.1.6 13-15 Fixed typo; missing section reference reference to 13.1.1 13.1.1 13.1.8 1345 13-'15 Changed "incredible" to "not Changed "incredible" credible" for clarification.
"not credible" 13.1.9 13.1.9 13-16,17 13-16,17 Updated reference reference to Figure 2-122412 (was Figure 2.10) and deleted reference reference to drain drain thimble used for cooling cooling tower blowdown (no longer exists). Also corrected typo referring referring to case 2 when it should be case 3. 3.
13.2/3 13-17 Updated section number from 13.3 to 13.2 and andfrom from 13.4 to 13.3 because because ofof combination combination of sections 13 13.1/2 13.1.
.112 into 13.1.
13.3 13.3 13-18 Added new reference 9 for the pool drain equation used in section 13.1.3.1 Chapter 14 All All outline numbering was revised to include a "TS" "TS" Prefix to avoid confusion with earlier chapters of the SAR. 1.1 1.1 14-1 Updated summary to reflect elimination of Rev. 0 redlining. 1.3.2 14-5 Added new definition for non-secured experiment Gust non-secured experiment (just opposite of secured secured experiment which was already experiment already defined). , 2.2 14-10,11 14-10,11 Clarified between first and second LSSS in basis. 3.1.2 - 14-13,14 14-13,14 Corrected control control rod to control control element. 3.2.5 3.2.5 14-22 Clarified wording wording for interlocks preventing application of air to fire transient preventing application transient rod. 3.4 14-26 Renamed Renamed section heading to "Confinement."
"Confinement." Response to NRC
#11I (I) comment #11
SAR Revision 2 changes changes Page 12 of 14 Section Rev2 Page Change Reference Reference 3.4 14-27 Updated stack height from 17m to 26.5m, and corresponding corresponding reduction fraction from from 10 to 2.6. 3.7.2 14-29 Updated release release concentration concentration and dilution fraction for new vent system. RSC 879 (9) lV) 3.8.1 14-30 Clarified limits on experiment experiment reactivity reactivity are fora for. a single experiment. 3.8.2 14-31 14-31 Revised outline outline numbering numbering for consistency. 3.8.2 14-31 14-31 Updated dilution fraction for new vent system. (9) RSC 879 lV ) 3.8.2 14-32 14-32 Changed formatting of basis list for clarification clarification only. 3.8.3 14-32 14-32 Clarified that the experimental capsule would be removed and inspected inspected if failure occurs (fuel may be inspected inspected ifif warranted, but would not be removed). 4.2 14-36 Clarified wording. Response to NRC comment # 14 (I) { 4.2 14-37 Updated Updated 40 year history to 50 years. 4.4 14-38 Renamed Renamed section heading heading to "Confinement."
"Confinement." Response to NRC comment #11
#11 (\))
4.5 14-39 39 Revised the basis for quarterly quarterly checks on ventilation ventilation system. 5.1 14-41 Changed stack height from 17m to 26.5m. (9) RSC 879 lV) 5.2 14-41 14-41 Deleted outlet pipe from 15 foot requirement requirement (outlet pipe requirement is specified specified in the next sentence). sentence). 5.3 14-42 Revised outline Revised outline numbering numbering for consistency. 6.1.1 14~45 14-45 Added reference reference to new Figure Figure 14-1 (just copy of Figure 12-1).12-1). 6.1.2 14-45 Fixed gender language. 14-1 Figure 14-1 14-46 Added new figure for clarification (copy of Figure 12-1). 12-1). 6.1.3 14-47 Clarified wording in sub-section sub-section 1-b. I-b. 6.2.1 14-48 Revised list of required required qualifications qualifications for RSC. 6.3 14-49 Clarified Clarified authority authority of state license and updated license license number. 6.4 14-50 Fixed gender language. language. 6.5(2) 14-51 Fixed gender language and clarified wording. 6.6.2 14-52 Fixed gender language. Fixed. 6.7.1 14-53,54. 14-53,54 Reordered Reordered sub-sections sub-sections under item 1 and revised outline numbering for consistency.
SAR Revision 22 changes SAR Revision changes Page 13 Page 13 of of 14 14 Section Section Rev2 Page Rev2 P~ge Change ChanKe Reference Reference Appendix A Appendix A headings and Revised headings Revised order of and order of entire Appendix to entire Appendix make itit clear to make sections of which sections clear which of the the SAR the calculations are supporting. SAR the calculations are supporting. A-i A-I Updated stack
.Updated stack height 17.1 to from 17.1 height from to 26.5m.
26.5m. RSC RSC 879 (9) . 879l~) A-i A-I equation 22 to Updated equation Updated reflect new to reflect stack height. new stack height. RSC 879 (9) RSC 879l~) 2) A-2 A-2 Updated equation 5 to reflect new building cross-section Updated equation 5 to reflect new building cross-section (12,200ft (12,200ftL) RSC 879 RSC (9) 879l~) A-2 A-2 equation 66 to Updated equation Updated to reflect changes in reflect changes equation 5. in equation 5. RSC 879 RSC 879l~) (9) A-2 A-2 Added number Added for equation 7 (previously number for equation 7 (previously not numbered).not numbered). A-2 A-2 number for Added number Added (previously not equation 88 (previously for equation not numbered). numbered). A-2 A-2 Added number Added for equation 9 (previously number for equation 9 (previously not numbered).not numbered). A-3 A-3 Added number for Added number equation .10 for equation (previously not J 0 (previously not numbered). numbered).
. A-3 A-3 number for Added number Added equation 11 for equation (previously not 11 (previously not numbered).
numbered) . A-3 A-3 Renumbered equation Renumbered equation 12, equation 99 to equation 12, and and updated equation to updated equation reflect flow-rate to reflect flow-rate RSC 879 RSC (9) 879l~) changing from 1000 changing 1000 scfmscfm toto 2700 scfm (also 2700 scfm (also clarified the unit clarified the unit conversion). conversion). A-3 A-3 equation 10 Renumbered equation Renumbered 10 to equation 13, to equation 13, and updated equation.to and updated changes in reflect changes equation.to reflect in RSC RSC 879 (9) 879l~) equations 22 and 12. equations 12. A-4 A-A equation 11 Renumbered equation Renumbered 11 to equation 14, to equation 14, and equation to updated equation and updated to reflect changes in reflect changes in RSC 879 RSC (9) 879l~) equations 6 and 12. equations 12. A-4 Updated factor from 10.17 Updated 10.17 to to 2.6 because of 2.6 because vent system. new vent of new system. RSC 879 (9) 879l~) r- A-4 A-4 Deleted and replaced sentence referring to building wake effects Deleted and replaced sentence referring to building wake effects vs. reactor lab vs. reactor lab RSC 879 (9) 879l~) wake effects being a 10 fold difference. From now on all wake effects are wake effects being a 10 fold difference. From now on all wake effects are calculated using the calculated using new ME the new ME building (12,200 ft2) dimensions (12,200 building dimensions ft2) since reactor lab the reactor since the lab is entirely enclosed entirely enclosed within the ME building. A-4 Clarified intermediate Clarified calculation of intermediate calculation of stack stack outlet concentration, and outlet concentration, clarified that and clarified that equation 15 equation 15 (which uses equation 2) is assuming uses equation 2) assuming stack height.stack height. A-4 A-4 Renumbered equation 77 to Renumbered equation equation 15, to equation 15, and and updated equation to updated equation changes in reflect changes to reflect in RSC 879 879l~) (9) equation equation 2.2.
SARRevision SAR Revision 22 changes changes Page Page 1414 of 14 Section Section Rev2 Page Rev2 Page Change Change Reference Reference A-4 Deleted calculation Deleted calculation of discharge discharge concentration concentration using using only reactor reactor lab lab building building cross-cross- RSC 879 (9) (lJ) . section (in support section support of of section section 11.1.1.1), 11.1.1.1), now only calculates calculates using entire ME ME building cross-section building cross-section which appropriate. which is more appropriate. A-4 . Clarified Clarified that height with that equation 16 is for zero stack height with building building wake effect. RSC (9) 879 (lJ) A-4 Renumbered equation Renumbered equation 9 to equation equation 16, 16, and updated updated to reflect reflect changes changes in in equation equation 6. RSC 879 (lJ) (9) A-4 Updated final sentence Updated sentence because now only 2 calculations are done, because now done, not 3 (because the (because the RSC (9) 879 (lJ) reactor-lab-only wake reactor-lab-only wake effect effect calculation calculation was removed). AlsoAlso updated reference reference to equation number equation number 1515 being more more realistic, but added added that the more more conservative conservative 16 was used in text. equation 16 equation . References References (1) 812, "NRC Request for Additional Information", RSC 812, Information", May May 6, 2004. (2) (2) 773, "Reactor RSC 773, Recorder Replacement",
"Reactor Pulse Channel Digital Recorder March 3, Replacement", March 3,2003.
2003. (3) (3) 803, "Replace Console RSC 803, "Replace Console Recorder with Honeywell Multitrend Paperless Recorder", March Multitrend Paperless Recorder", March 12, 12,2004. 2004. (4) (4) RSC 887, "Install New Neutron Measuring "Install New Measuring Channels; Picoammeters", May 1,2006. Channels; LCR, LogN, and Picoammeters", 1, 2006. (5) (5) "Installation of New 893, "Installation RSC 893, New Building Building Reactor Reactor Evacuation Alarm System", Evacuation Alarm August 4, System", August 2006. 4,2006. (6) (6) "Installation of New 896, "Installation RSC 896, New Reactor Reactor Stack Stack Air and Continuous Continuous Air Monitors", Monitors", August 4, 4,2006. 2006. (7) (7) Alternating Current SCRAM", October 31, RSC 856, "Loss of Alternating 31, 2005. (8) (8) RSC 857, "Pneumatic "Pneumatic Tube Replacement", OCtober 31,2005. October 31, 2005. (9) "Modification to Reactor Ventilation System", March RSC 879, "Modification 21, 2006. March 21,2006. (10) (10) RSC 895, "Installation of Un 895, "Installation interruptible Power Supply", Uninterruptible Supply", August 4,2006. 4, 2006. (11) (11) RSC 869, "Report 2005-2006 Annual Maintenance "Report on 2005-2006 Activities", February Maintenance Activities", 14, 2006. February 14,2006. (12) (12) RSC 919, "Report 2006-2007 Annual Maintenance "Report on 2006-2007 Maintenance Activities", Activities", February February 1,2007. 1, 2007. (13) (13) RSG 894, "Installation RSC "Installation of New Fire Detection and Suppression System", August 4, 2006. (14) (14) RSC RSC Charter, as revised on December 6, 2006. See also RSCM-72, RSCM-72; "Minutes "Minutes of the UWNR Safety Committee Meeting, Meeting, December 6, 2006". (15) (15) RSC 922, "Dosimetry "Dosimetry for Students in NE 234, 427 and 428", March 14,2007. 14, 2007. r
- ANALYSIS REPORT SAFETY ANALYSIS FOR RENEWAL OF LICENSE R-74 FOR THE UNIVERSITY UNIVERSITY OF WISCONSIN NUCLEAR REACTOR NUCLEAR REACTOR
- Rev 2, 2, September September 2008
- REVISION REVISION HISTORY Rev 2, September 2008 2008 changes related to the remodeling This revision primarily consists of changes remodeling of the Mechanical Mechanical Engineering Building. New floor plans and room layouts were described. New equipment equipment installed included the ventilation system, demineralizer, and the pneumatic tube system. The console instrumentation significantly upgraded. Because this revision impacted so much of instrumentation was significantly of document, the opportunity was taken to make many minor corrections the document, corrections and clarifications. As a result, this revision was issued as a complete replacement to Rev 1, complete replacement 1, since over 95%
95% of pages removed and will no longer be maintained. All changes from Rev were updated. The index was removed 1 are indicated with a vertical line in the margin for each each line changed. Rev 1, August 2004 The reasons for this revision were a new reactor pool water cooling system, and a new computer-computer-channel. Chapters 5 and 7 were completely revised, and the Table of Contents and based pulse channel. changes from Rev 0 were indicated with a completely revised accordingly. All changes Index were completely vertical line in the margin for each line changed. Change indications in Rev 0 (Chapter (Chapter 14) have been removed. 0, April 2000 Rev 0,
- This was a major revision issued as part of the application for license renewal in 2000. Sections NUREG-153 7, but no substantial reorganized to conform with the format guidelines in NUREG-1537, were reorganized changes were made to the content of the SAR with the exception of the Technical Technical Specifications (Chapter changes were indicated with a vertical line in the margin for each line 14). These changes (Chapter 14).
changed. changed .
~THE . .................................................. 1-1 FACILITY.................................................
- 1 1 THE FACILITY 1-1 1.1 1.1 Introduction. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 1-1 Introduction................................................ 1-1 1.2 1.2 Summary and Summary and Conclusions Conclusions on Principal.Safety on Principal Safety Considerations Considerations ....... ....... 1-1 1-1 General Description 1.3 General Description of of the the Facility Facility..............................
............................... 1-2 1-2 1.4 1.4 Shared Facilities Shared Facilities and and Equipment ...............................
Equipment ................................ 1-6 1-6 1.5 1.5 Comparison With Comparison With Similar Similar Facilities Facilities ..............................
............................. 1-6 1-6 1.6 1.6 Summary of Summary of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 1-6 Operations ....................................... 1-6 1.7 Compliance With 1.7 Compliance With the the Nuclear Nuclear Waste Waste PolicyPolicy Act Act of 1982..............1-7 of 1982 ............. 1-7 1.8 Facility Modifications 1.8 Facility Modifications and and History..............................
History ............................... 1-8 1-8 22 SITE CHARACTERISTICS ....................................... SITE CHARACTERISTICS ......................................... 2-1 2-1 2.1 2.1 Geography and Geography Demography .................................. and Demography . ................................... 2-1 2-1
- 2. 1.1 Site Location and Description...............................
2.1.1 Description ................................ 2-1 2-1
- 2. 1.1.1 Specification 2.1.1.1 Specification and Location Location............................
............................. 2-1 2-1 2.1.1.2 Boundary Boundary and Zone ........................
Zone Area Maps ......................... 2-1 Population Distribution.................................... 2.1.2 Population Distribution ..................................... 2-2 2.2 2.2 Nearby Nearby Industrial, Transportation, and Industrial, Transportation, and Military Military Facilities Facilities ..........
. ......... 2-14 2-14 2.2.1 Locations ....................................
Locations and Routes ...................................... 2-14 2-14
............................................
2.2.2 Air Traffic .............................................. 2-15 2-15 2.2.3 Analysis of Potential 2.2.3 Accidents at Facilities .................... Potential Accidents .................... 2-15 2-15 2.3 2.3 Meteorology............................................... Meteorology . ................................................ 2-15 2-15 2.3.1 General and Local Climate................................ 2.3.1 General Climate ................................. 2-15 2-15
- 2.3.2 2.3.2 2.4 Hydrology 2.4 Site Meteorology 2.3.2.2 Precipitation 2.3.2.2 2.3.2.3
.......................................
Meteorology ......................................... 2-16 2.3.2.1 Temperature 2.3.2.1 Temperature .....................................
....................................... 2-16 Precipitation .....................................
....................................... 2-16 2.3.2.3 Winds..........................................
Winds ............................................ 2-19 Hydrology................................................
. ................................................. 2-20 2-16 2-16 2-16 2-19 2-20 2.5 2.5 Geology, Geology, Seismology, Seismology, and Geotechnical Engineering and Geotechnical Engineering ...............
............... 2-23 2-23 2.5.1 2.5.1 Regional Regional Geology .......................................
Geology ......................................... 2-23 2-23 2.5.2 2.5.2 Site Geology........................................... Geology ............................................. 2-24 2-24 2.5.3 Seismicity............................................. 2.5.3 Seismicity ............................................... 2-24 2.5.4 Maximum Earthquake 2.5.4 Maximum Earthquake Potential ............................ Potential ............................. 2-25 2-25 2.5.5 2.5.5 Vibratory Ground Motion................................. Vibratory Ground Motion .................................. 2-25 2-25 2.5.6 Surface 2.5.6 Surface Faulting ........................................ Faulting .......................................... 2-25 2-25 2.5.7 Liquefaction 2.5.7 Potential.................................... Liquefaction Potential ..................................... 2-25 2-25 2.6 2.6 References................................................ References . ................................................. 2-26 2-26 3 DESIGN DESIGN OF OF STRUCTURES, STRUCTURES, SYSTEMS, SYSTEMS, AND COMPONENTS...........3-1 AND COMPONENTS .......... 3-1 3.1 Design Criteria ............................................... Design Criteria ............................................. 3-1 3-1 3.2 3.2 Meteorological Meteorological Damage ...................................... Damage ........................................ 3-1 3-1 3.3 Water Damage.............................................. Water Damage . ............................................... 3-2 3-2 3.4 Seismic Damage............................................. 3.4 Seismic Damage . .............................................. 3-2 3-2 3.5 Systems and 3.5 Systems and Components ...................................... 3-2 Components ...................................... 3-2
- Report Rev. 2 UWNR Safety Analysis Report UWNR TOC-1 TC1Sp.20 TOC-l Sept. 2008 2008
44 3.6 3.6 4.1 R eferences ................................................... References REACTOR DESCRIPTION
. .................................................. 3-2
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1 DESCRIPTION .........................................
Description ........................... Summary Description 4.1.1 Introduction
. ......................................... 4-1 Introduction ...............................
.............................................. 4-1 4 -1
- 4.1.2 Summary Summary of Reactor Data ....................
................................... 4-3 4.1.3 Experimental Experimental Facilities ......................
Facilities ..................................... 4-5 4.1.3.1 Hydraulic Irradiation Irradiation Facility (Whale) .................... ..... 4-5 4.1.3.2 Thermal Thermal Column .....................
.................................... 4-6 4.1.3.3 Beam Ports ..........................
......................................... 4-6 4-6
..................................... 4-6 4.1.3.4 Pneumatic Tube ...................... 4-6 4.2 R eactor Core ..................................
Reactor . ................................................ 4-6 4-6 4.2.1 Reactor Reactor Fuel ..............................................
............................... ........ 4-8 4-8 4.2.2 Control Elements .........................................
........................... 4-13 4-13 I 4.2.2.1 Control 4.2.2.1 Control Blade Shrouds Shrouds and Guide Tubes Tubes... ................. 4-13 4-13 I 4.2.2.2 Safety Blades ........................
Blades ...................................... 4-14
- 4-14
_I 4.2.2.3 Regulating ..................... Regulating Blade ................................... 4-14 4-14 I 4.2.2.4 4.2.2.4 Transient Transient Control Rod ...............................
................. 4-15 4-15 4.2.3 4.2.3 Neutron Moderator and Reflector ..............
Reflector ............................ 4-16 4-16 4.2.4 Neutron Startup Source ....................................
...................... 4-17 4-17 4.2.5 Core Support Structure Structure ....................................
...................... 4-20 4-20 Reactor Pool ...................................
0 4.3 Reactor . ................................................ 4-24 4-24
- 4.4 ...............................
Biological Shield ............................................. 4-26 4-26 4.4.1 Pool Surface Radiation Levels - N16 Surface Radiation 16 Activity ...................
..... 4-27 4-27 4.4.2 Heating Effects in Shield and Thermal Column ................. ... 4-27 4-27 4.5 Nuclear Design ..............................................
................................ 4-27 4-27 4.5.1 Normal Normal Operating Operating Conditions and Reactor Core Physics Parameters Parameters. 4-27 4-27 4.5.1.1 Arrangements .................................
4.5.1.1 Core Arrangements ................................ 4-28 4-28 4.5.1.2 Standard TRIGA fuel cores ........................... 4.5.1.2 Standard .......................... 4-29 4-29 4.5.1.2.1 4.5.1.2.1 Reactivity Reactivity Effects Effects In StandardStandard Fuel Cores .... ... 4-32 4-32 4.5.1.3 Cores containing FLIP fuel ...........................
.......................... 4-34 4-34 4.5.1.3.1 First mixed core- 9 FLIP bundles and 16 standard bundles ....................
bundles ..................... 4-37 4.5.1.3.2 Second mixed core - 15 FLIP fuel bundles and 10 standard standard bundles ................ bundles ................. 4-40 4.5.1.3.3 ....................... All-FLIP core ......................... 4-44 Temperature Coefficient 4.5.1.4 Isothermal Temperature ................... Coefficient .................... 4-47 4.5.1.5 Pulse Parameters Param eters ..................................
................................... 4-48 4.5.2 Operating Lim its .........................................
Operating Limits ........................................ 4-51 4.6 Thermal-Hydraulic Thermal-Hydraulic Design ...................................
.................................... 4-52 4.7 R eferences ..................................................
References ................................................. 4-52 UWNR Safety Analysis Report Rev. 2 TOC-2 TOC-2 2008 Sept. 2008
- 0
- 55 REACTOR REACTOR COOLANTCOOLANT SYSTEMS..................................
SYSTEMS ................................... 5-1 5-1 5.1 5.1 Description........................................ Summary Description ......................................... 5-1 5-1 5.2 Primary 5.2 Coolant System...................................... Primary Coolant System ....................................... 5-1 5-1 5.3 Intermediate Coolant System .................................. 5.3 Intermediate Coolant System . ................................... 5-1 5-1 5.4 Campus 5.4 Campus Chilled Chilled Water System................................. Water System .................................. 5-3 5-3 5.5 Primary 5.5 Coolant Cleanup Primary Coolant Cleanup System System ...............................
. ....................... "........ 5-3 5-3 5.6 Primary Coolant Makeup Water System 5.6 Primary Coolant Makeup Water System .......................... 5-4......................... 5-4 5.7 Nitrogen-16 5.7 Control System Nitrogen-16 Control System ....................................
.................................... 5-4 5-4 5.8 5.8 Auxiliary Systems Auxiliary Systems Using Using Primary Primary Coolant Coolant ........................
. ........................ 5-4 5-4 5.9 5.9 References.................................................
References . .................................................. 5-8 5-8 66 ENGINEERED ENGINEERED SAFETY SAFETY FEATURES................................ FEATURES ................................. 6-1 6-1 6.1 6.1 Summary Description .......................................... Summary Description ........................................ 6-1 6-1 6.2 6.2 Detailed Descriptions .......................................... Detailed Descriptions ........................................ 6-1 6-1 6.2.1 Confinement............................................ 6.2.1 Confinement .............................................. 6-1 6-1 6.2.2 Containment............................................ 6.2.2 Containment .............................................. 6-2 6-2
~6.2.3 Emergency
*6.2.3 Emergency Core Cooling System .............................
Core Cooling ............................ 6-2 6-2 6.3 References................................................. 6.3 References . .................................................. 6-2 6-2 77 INSTRUMENTATION AND INSTRUMENTATION AND CONTROL CONTROL SYSTEMS.................... SySTEMS .................... 7-1 7-1 7.1 7.1 Summary Description Summary Description ........................................
. ......................................... 7-1 7-1 7.2 Design of of Instrumentation Instrumentation and and Control Control SystemsSystems ...................
. ................... 7-1
- 7.2 Design 7-1 7.2.1 Design Criteria 7.2.1 Criteria..........................................
............................................ 7-1 7-1 7.2.2 Design-Basis 7.2.2 Requirements.................................
Design-Basis Requirements .................................. 7-3 7-3 7.2.3 System Description 7.2.3 ........................................ 7-3 Description ...................................... 7-3 7.2.3.1 Start-up Channel................................... 7.2.3.1 Start-up Channel .................................... 7-3 7-3 7.2.3.2 7.2.3.2 Log N Channel.............................. N - Period Channel ............................... 7-3 7-3 7.2.3.3 7.2.3.3 Pulse Power Channel................................ Power Channel ................................. 7-3 7-3 7.2.3.4 Safety Channels.................................... 7.2.3.4 Safety Channels ..................................... 7-4 7-4 7.2.3.5 Temperature Measurements........................... 7.2.3.5 Temperature Measurements ............................ 7-4 7-4 Performance Analysis............................... 7.2.4 System Performance Analysis ................................ 7-4 7-4 7.2.5 Conclusion............................................. 7.2.5 Conclusion ............................................... 7-4 7-4 7.3 Reactor Control System ...................................... 7.3 Reactor Control System ........................................ 7-5 7-5 7.3.1 7.3.1 Mode ........................................... Mode Switch ............................................. 7-5 7-5 7.3.2 Manual 7.3.2 Manual Operation ....................................... Operation ......................................... 7-5 7-5 7.3.3 Square 7.3.3 Square Wave Operation ................................... Operation ..................................... 7-5 7-5
........................................
Operation .......................................... 7.3.4 Pulsing Operation 7-6 7-6 7.3.5 Operation .................................. 7-8 7.3.5 Control Element Operation................................. 7-8 7.3.6 Safety Blade 7.3.6 Control...................................... Blade Control ....................................... 7-8 7-8 7.3.7 Regulating Blade 7.3.7 .................................. Control .................................... Blade Control 7-8 7-8 7.3.8 Transient 7.3.8 Transient Rod Control ....................................
...................................... 7-9 7-9 7.3.9 Automatic 7.3.9 Automatic Level Control System.............................
System .............................. 7-9 7-9 7.4 Reactor Protection 7.4 Reactor Protection System System (Scram (Scram Circuits) Circuits) .....................
...................... 7-10 7-10
- Analysis Report Rev. 2 UWNR Safety Analysis TOC-3 TOC-3 2008 Sept. 2008
7.5 7.5 7.6 7.6 Engineered Control 7.6.1 7.6.2 Safety Features Engineered Safety Control Console Console and 7.6.1 Alarm 7.6.2 Indicator 7.6.3 Features Actuation and Display Actuation Systems Display Instruments Systems .................... System................................ Indicator System Alarm and Indicator
...................
Instruments ........................
.......................
................................. 7-13
........................................
Indicator Lights .......................................... Pneumatic Tube System Panel.............................. 7.6.3 Pneumatic Panel ............................... 7-14 7-13 7-13 7-13 7-13 7-13 7-14 7-14 7-14
- Ventilation System Panel..................................
7.6.4 Ventilation Panel ................................... 7-15 7-15 7.6.5 7.6.5 Cooling Cooling System System PanelPanel....................................
..................................... 7-15 7-15 7.6.6 7.6.6 Whale .....................................
Whale System Panel ....................................... 7-15 7-15 7.7 7.7 Radiation Radiation Monitoring Monitoring SystemsSystems................................
................................. 7-15 7-15 7.7.1 Area Radiation 7.7.1 Radiation Monitors..................................
Monitors ................................... 7-15 7-15 7.7.2 Stack Air Monitor 7.7.2 ...................................... Monitor ........................................ 7-16 7-16 7.8 7.8 References................................................ References. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-17 7-17 88 ELECTRICAL POWER ELECTRICAL POWER SYSTEMS SYSTEMS ...................................
................................. 8-1 8-1 8.1 8.1 Normal Electrical Normal Electrical Power Power Systems Systems...............................
................................ 8-1 8-1 8.1.1 277/480 VAC 8.1.1277/480 VAC 33 Phase ElectricalElectrical Power ........................
........................ 8-1 8-1 8.1.2 277/480 VAC 8.1.2277/480 VAC 33 Phase Backed Up Electrical Power Up Electrical Power ...............
............... 8-1 8-1 8.1.3 240 8.1.3 2404 4 wire VAC Electrical VAC Electrical Power, Transformer Transformer T T1I................8-2
............... 8-2 8.1.4 208/120 3 wire VAC 8.1.420811203 Electrical Power, VAC Electrical Transformer T2 Power, Transformer T2............8-2
........... 8-2 8.1.5 277 VAC 8.1.5 277 VAC Lighting Electrical ..........................
Electrical Power ........................... 8-3 8-3 8.1.6 277 V 8.1.6 277 VAC AC Backed Backed Up Up Lighting ~lectrical Electrical Power..................8-3 Power ................. 8-3 8.2
- 8.2 Emergency Electrical Emergency Electrical Power Power Systems Systems............................
............................. 8-3 8-3 8.2.1 Reactor Laboratory 8.2.1 Reactor Uninterruptible Power Supply Laboratory Uninterruptible Supply .... I............8-3
................. 8-3 8.2.2 Reactor Laboratory 8.2.2 Reactor Laboratory SupportSupport Space Space Uninterruptible Uninterruptible Power Supply .... ... 8-4 8-4 8.2.3 Mechanical 8.2.3 Engineering Building Mechanical Engineering Building Emergency Emergency Power Power..............8-4
............. 8-4 99 AUXILIARY AUXILIARY SYSTEMSSYSTEMS ..........................................
............................................ 9-1 9-1 9.1 9.1 Heating, Ventilation, and Heating; Ventilation, and Air Conditioning Systems Air Conditioning Systems................
................ 9-1 9-1
- 9. 1.1 Air Handling Unit 9.1.1 Air Handling Unit .......................................
........... .' ............................. 9-1 9-1 9.1.2 Exhaust 9.1.2 Fans ...........................................
Exhaust Fans ............................................. 9-1 9-1 9.1.3 Filters .................................................. 9.1.3 ................................................ 9-2 9-2 Variable Air Volume Boxes 9.1.4 Exhaust Variable Boxes .........................
.......................... 9-2 9-2 9.1.5 Emergency Venting Mode .................................
9.1.5 Emergency Venting Mode .................................. 9-3 9-3 9.1.6 9.1.6 Beam Beam Port Thermal Column Ventilation System .............. Port and Thermal .............. 9-3 9-3 9.1.7 9.1. 7 Pneumatic Pneumatic System System Fume Hood Exhaust Exhaust........................
........................ 9-4 9-4 9.2 9.2 Handling Handling andand Storage Storage of of Reactor Reactor Fuel Fuel...........................
............................ 9-8 9-8 9.2.1 9.2.1 Fuel Handling...........................................
Handling. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-8 9-8 9.2.2 Fuel Storage 9.2.2 Storage............................................
.............................................. 9-9 9-9 9.2.3 Fuel Bundle Maintenance 9.2.3 Maintenance and Measurements ...................
Measurements .................... 9-9 9-9 9.3 Fire Protection 9.3 Fire Protection Systems and Programs.......................... and Programs ........................... 9-13 9-13 9.4 9.4 Communication Systems Communication Systems.....................................
...................................... 9-13 9-13 9.5 9.5 Possession and and Use of of Byproduct, Byproduct, Source, Source, and and Special Nuclear Material Special Nuclear Material......................................
...................................... 9-13 9-13 UWNR Safety Analysis Report Rev. 2 TC4Sp.20 TOC-4 TOC-4 2008 Sept. 2008
- Primary Coolant Coolant Systems Systems ............. 9-14
- 9.6 9.6 Cover Gas Cover Gas Control Control in in closed closed Primary ............. 9-14 9.7 9.7 Other Auxiliary Other Auxiliary Systems.....................................
Systems ...................................... 9-14 9-14 9.8 9.8 References................................................ References .................................................. 9-14 9-14 10 10 EXPERIMENTAL FACILITIES EXPERIMENTAL FACILITIES AND AND UTILIZATION UTILIZATION .................. .................. 10-1 10-1 10.1 Summary Description 10.1 Summary ....................................... Description ......................................... 10-1 10-1 10.2 Experimental Facilities 10.2 Experimental ...................................... Facilities ........................................ 10-1 10-1 10.2.1 Thermal Column Column ...................................
........................... 10-1 10-1 10.2.2 10.2.2 ......................................
Beam Ports ........................................ 10-3 10-3 10.2.3 10.2.3 Pneumatic Tube Pneumatic Tube...................................
.................................... 10-6 10-6 10.2.4 10.2.4 Grid Box Irradiation . . . . . . . . . . . . . . . . . . . . . . .. 10-8 Irradiation Facilities ......................... 10-8 10.3 Experiment 10.3 ........................................
Review .......................................... Experiment Review 10-11 10-11 10.4 References............................................... 10.4 References ................................................. 10-12 10-12 11 11 PROTECTION PROGRAM RADIATION PROTECTION RADIATION PROGRAM AND AND WASTE MANAGEMENT WASTE MANAGEMENT........................................
.......................................... 11-1 11-1 11.1 Radiation 11.1 Protection........................................
Radiation Protection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 11-1 11-1 11.1.1 11.1.1 Sources ................................ Radiation Sources Radiation .................................. 11-1 11-1 11.1.1.1 Radiation Sources.........................
- 11. 1.1.1 Airborne Radiation Sources . . . . . . . . . . . . . . . . . . . . . . . . .. 11-1 11-1 11.1.1.1.1 11.1.1.1.1 Releases from abnormal abnormal reactorreactor operations..
operations . .. 11-1 11-1 11.1.1.1.2 Releases from normal reactor operations normal reactor operations ... . . . .. 11-2 11-2
- 11. 1.1.2 Liquid 11.1.1.2 Radioactive Sources ................
Liquid Radioactive . . . . . . . . . . . . . . . . . . . . . . . . ...11-4 11-4 .
- 11.1.2 11.1.3 11.1.3 11.1.4 11.1.4 11.1.5 11.1.S 11.1.1.3 11.1.1.3 Solid Radiation ALARA Radiation Radioactive Sources .................
Radioactive Protection Program Radiation Protection Program ............... ALARA Program.........................
......................... l1-S Program ................................. " 11-S Monitoring and Surveying ..........
Radiation Monitoring Radiation Exposure Control and Dosimetry Radiation
................' .... l1-S Dosimetry ..... .............. 11-S 11-4
. . . . . . . . . . . . . . . . . . . . . . . . . .. 11-4 11-5 11-5 11-5 11-5 11.1.6 11.1.6 Contamination Control Contamination ....................
Control .............................. 11-6 11-6 11.1.7 11.1.7 Monitoring.................. Environmental Monitoring. Environmental 11-6
. . . . . . . . . . . . . . . . . . . . . . . . . .. 11-6 11.2 Radioactive 11.2 Radioactive Waste Waste Management Management ......................
................................ 11-6 11-6 11.2.1 Radioactive Waste Management Radioactive Management Program Program ....... ................ 11-6 11-6 11.2.2 Radioactive Waste Control Radioactive Control .................
........................... 11-7 11-7 11.2.3 11.2.3 Radioactive Waste .............
Release of Radioactive Release ......................... 11-7 11-7 11.3 References 11.3 References 11-7 11-7 12 12 CONDUCT OF CONDUCT OF OPERATIONS .................................... OPERATIONS ...................................... 12-1 12-1 12.1 Organization 12.1 Organization ..............................................
................................................ 12-1 12-1 12.1.1 Structure .........................................
Structure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 12-1 12-1 12.1.2 Responsibility.................................... Responsibility. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 12-1 12-1 12.1.2.1 Radiation Safety University Radiation Safety Committee Committee............12-1
........... 12-1 12.1.2.2 12.1.2.2 Radiation Safety University Radiation Safety (Part of University Department Department of Environment, Health and Safety) ............... ............... 12-1 12-1 12.1.2.3 12.1.2.3 Engineering Physics Department Chair, Engineering Department ...................... 12-1
- UWNR Safety Analysis Report Rev. 2 TOC-5 TC5Sp.20 TOC-S 2008 Sept. 2008
12.1.2.4 12.1.2.4 12.1.2.5 12.1.2.5 12.1.2.6 12.1.2.6 12.1.2.7 12.1.2.7 12.1.2.8 12.1.2.8 Reactor Director .............................. Senior Operators Senior
..........
Reactor Supervisor ............................ Operators (alternate Reactor operators
........
. . . . . . . . . . . . . . . . . o. ..
Reactor Safety Committee ... . . . . . . . . . . . . . . . . . . . .. 12-2
.................
.......ors) ...........
(alternate Supervisors) Su operators .............................
.........
.........
........... 12-3
)........
o.. o... 12-2 12-2 12-2 12-2 12-2 12-3 12-3 12-3
- 12.1.3 Staffi ng .......................
Staffing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 12-4
. . . . . . . . . . . . . . . . . .o °.. 12-4 12.1.4 12.1.4 Selection and Training of Personnel . . . . . . . . . . . . . . . . . ,. .. 12-4 Personnel .................... 12-4 12.1.5 12.1.5 Radiation ................
Radiation Safety .................................... . . . . . . . . . . . . . . . . . °.. ° 12-5 12-5 12.2 Review Review and Audit Activities Activities .................
. . . . . . . . . . . . . . . .. .. .. .. .. .. .. .. .. .. .. .. . .. .. .. .. .,..... 12-7 12-7 12.2.1 Composition and Qualifications Qualifications .......................
... 12-7 12-7
. . . . . . . . . . . . . . . . . . . .° 12.2.2 Charter and Rules ................ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 12-7 12-7
.................. ,..
12.2.3 Review Function ...................................
............... 12-7 12-7
. . . . . . . . . . . . . . . . . .°. .
12.2.4 ................. Audit Function ..................................... 12-8 12-8
. . . . . . . . .°. . . . . . .,, . , .
12.3 12.3 .......................... Procedures .................................................. 12-8 12-8
....................
12.4 12.4 Required Actions .........................
. ............................................ 12-9 12-9
................. o..
12.5 R eports .................................................... Reports ................................. 12-10 [2-10 12.6 12.6 R ecords ................................. Records .................................................... ................. [2-13
.. 12-13 12.7 12.7 Emergency Emergency Planning ......................
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 12-14
................. .. 12-14 12.8 12.8 Security Planning ........
Security ........ *...............
~ . . . . . . . . . . . . . . . .................
. . . . . . . . . . . . . . . . . .. °.. [2-15 12-15 12.9 12.9 Quality Assurance ........................
Quality . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 12-15 12-15
................... ° 12.10 Operator Training and Requalification Operator Requalification ...... ......................... 12-16 12-16 Startup Plan ..............................
0 12.11 ............................................... 12-16 12-16 12.12 Envir~mmental Reports Environmental Reports ...................
...................................... 12-16 12-16
- 12.13 12.13 R ..............................
eferences ....................................... References ; ......... 12-16 12-16 13 13 ACCIDENT ACCIDENT ANALYSIS ANALYSIS ...........................................
........................................... 13-1 13-1 13.1 13.1 Accident Analysis Initiating Events and Determination Determination of of C onsequences ................................................
Consequences. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 13-1 13-1 13.1.1 13 .1.1 Hypothetical Accident ;..................... Maximum Hypothetical ...................... 13-1 13-1 13.1.1.1 Fission Product Inventory Inventory in Fuel Element Element ......... ......... 13-2 13-2 13.1.1.2 13.1.1.2 Fission Product Release Release Fraction ................. Fraction ................. 13-2 13-2 13.1.1.3 13.1.1.3 Activity in Pool Water W ater .........................
......................... 13-2 13-2 13.1.1.4 13.1.1.4 Fission Product Release Release to Air within the Reactor Reactor Laboratory ..................................
Laboratory .................................. 13-3 13-3 13.1.1.5 13 .1.1.5 Release Release of Fission Products to Unrestricted Areas Areas ... ... 13-6 13-6 13.1.2 Insertion of Excess Reactivity Reactivity .........................
......................... 13-8 13-8 13.1.2.1 Fuel Temperatures Temperatures from Operation Operation at the Scram Point 13-9 13-9 13.1.2.2 Temperature after Pulse .......................
TemperatureafierPulse ....................... 13-10 13-10 13.1.3 Loss of Coolant ...................................
................................... 13-11 13-11 13.1.3.1 13 .1.3.1 Possible Means of Water Loss ...................
. . . . . . . . . . . . . . . .. 13-11 13-11 13.1.3.2 13.1.3.2 Radiation Radiation Levels Levels in Confinement Confinement Due to Unshielded C ore ......................................
Core ...................................... 13-12 13-12 UWNR Safety Analysis Report Rev. 2 TOC-6 TOC-6 Sept. 2008
*
- 13.1.3.3 13.1.3.3 Radiation Radiation Levels Levels in Unrestricted Unrestricted Areas Areas Due to Unshielded .........................
Unshielded Core ............................. 13-12 13-12 13.1.3.4 13.1.3.4 Temperature After Loss of Pool Fuel Temperature Pool Water ... ....... 13-14 13-14 13.1.4 Loss of Coolant Flow .................... Coolant Flow. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 13-14 13-14 13.1.5 13.1.5 Mishandling or Mishandling or Malfunction Malfunction of Fuel Fuel .........
................... 13-14 13-14 13.1.6 13.1.6 Experiment Malfunction .................
Experiment Malfunction ............................ 13-15 13-15 13.1.7 13.1.7 Electrical Power ........... Loss of Normal Electrical ...................... 13-15 13-15 13.1.8 13.1.8 External Events ........................ External ................................... 13-15 13-15 13.1.9 13.1.9 Mishandling Malfunction of Equipment Mishandling or Malfunction Equipment ..............
.. 13-16 13-16 13.2 Summary 13.2 Summary and and Conclusions ........................
Conclusions .................................... 13-17 13-17 13.3 References..................................... 13.3 References ................................................. 13-17 13-17 14 14 TECHNICAL SPECIFICATIONS TECHNICAL ................................. SPECIFICATIONS ................................... 14-1 TS 11 INTRODUCTION TS INTRODUCTION ......................................
......................................... ; ..... 14-1 I TS 1.1 TS 1.1 Scope..........................................
Scope ............................................ 14-1 14-1 I TS 1.2 TS 1.2 Format......................................... Format ........................................... 14-1 14-1 I TS 1.3 TS 1.3 Definitions...................................... Definitions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 14-1 14-1 I TS 1.3.1 TS 1.3.1 Reactor Operating Conditions .................. Reactor Operating ................... 14-2 14-2 I TS 1.3.2 TS 1.3.2 Reactor Experiments Reactor Experiments and Irradiation Irradiation .............
.............. 14-4 14-4 I TS 1.3.3 TS 1.3.3 Components .........................
Reactor Components Reactor .......................... 14-5 14-5 I TS TS 1.3.4
1.3.4 Instrumentation
..................... Reactor Instrumentation: Reactor ...................... 14-7 14-7 I TS TS 22 SAFETY SAFETY LIMITSLIMITS AND LIMITING SAFETY AND LIMITING SAFETY SYSTEM SYSTEM SETTINGS. SETTINGS . 14-9 14-9 I
- TS TS TS 2.1 2.1 TS 2.2 TS 2.2 TS 33 LIMITING Safety Limits Safety Limiting LIMITING CONDITIONS TS 3.1 TS 3.1 TS 3.1.1 TS 3.1.1 Limits.......................
Limiting Safety Reactor Core Reactor Safety System CONDITIONS FOR FOR OPERATION Settings.................... System Settings Parameters ......................... Core Parameters
.................. 14-13
.......................... 14-13 Reactivity ...........................
Excess Reactivity 14-9
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 14-9 14-10
. . . . . . . . . . . . . . . . . . . .. 14-10 OPERATION ................. 14-13 14-13 14-13
. . . . . . . . . . . . . . . . . . . . . . . . . .. 14-13 I
I I I I TS 3.1.2 TS 3.1.2 Shutdown Margin Shutdown Margin........................... 14-13
. . . . . . . . . . . . . . . . . . . . . . . . . .. 14-13 I TS 3.1.3 TS 3.1.3 Pulse Limits .............................. 14-14
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 14-14 I TS 3.1.4 TS Core Configurations.........................
Configurations .......................... 14-15 14-15 I TS TS 3.1.5 3.1.5 Reactivity Coefficients....................... Reactivity Coefficients ........................ 14-16 14-16 I TS TS 3.1.6 3.1.6 Fuel Parameters............................ Parameters ............................. 14-17 14-17 I TS 3.2 TS 3.2 Reactor Control and Reactor Control and Safety Safety Systems Systems ................
................. 14-18 14-18 I TS 3.2.1 TS 3.2.1 Operable Control Operable ......................
Control Rods ....................... 14-18 14-18 I TS TS 3.2.2 3.2.2 Reactivity Insertion Rates (Scram Reactivity Insertion (Scram time)time)..........
.......... 14-19 14-19 I TS 3.2.3 TS 3.2.3 Limitations.............
Operation Limitations Other Pulsed Operation ............. 14-19 14-19 I TS 3.2.4 TS Reactor Safety System ............ Reactor Safety System ........................ ........... 14-20 14-20 I Table 3.2.4 Reactor Reactor Safety System Channels ............ ........... 14-20 14-20 I TS 3.2.5 TS 3.2.5 Interlocks..................... Interlocks .................................. ........... 14-22 14-22 I Table 3.2.5 Interlocks.................. 3.2.5 Interlocks ............................... ........... 14-22 14-22 I TS TS 3.2.6 3.2.6 Shutdown Mechanisms Backup Shutdown Mechanisms ... ........... 14-23
................. 14-23 I TS 3.2.7 TS 3.2.7 Bypassing Channels .............
Bypassing .......................... ........... 14-23 14-23 I
- Safety Analysis UWNR Safety Analysis Report Rev. 2 TOC-7 TC7Sp.20 TOC-7 2008 Sept. 2008
TS 3.2.8 TS 3.3 Control Systems and Instrumentation Instrumentation Required for Operation ...............................
...........................
Table 3.2.8 Instrumentation and Controls Required for O peration ........................................ Operation .................................... Reactor Pool Water Systems Systems ........................
....................
Required
.... 14-24 14-24 14-24 14-24 14-25
- 0 TS 3.4 Confinement .................................
Confinement . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 14-26 14-26 TS 3.5 Ventilation Systems Systems ...............................
........................... 14-27 14-27 TS 3.6 Emergency Power ................................
Emergency Power ............................ 14-28 TS 3.7 Radiation Monitoring Systems Systems and Effluents Effluents ....... . . . . . . . .. 14-28 TS 3.7.1 Monitoring Monitoring Systems ......................
.......................... 14-28 14-28 Table 3.7.1 Radiation Radiation Monitoring Systems .............. .......... 14-28 TS 3.7.2 Effluent (Argon-41)
(Argon-4l) Discharge Discharge Limit ............ ........ 14-29 14-29 TS 3.8 Experim ents ..................................... Experiments ................................. 14-30 14-30 TS 3.8.1 Reactivity Reactivity Limits ............................
........................ 14-30 14-30 TS 3.8.2 M aterials ..................................
Materials .............................. 14-31 TS 3.8.3 Experiment Failure and Malfunctions Experiment Malfunctions ............ ........ 14-32 14-32 TS 3.9 Facilitv SDecific Facility Specific LCOs .........................
............................. 14-32 14-32 TS 4 SURVEILLANCE REQUIREMENTS ..........................
SURVEILLANCE REQUIREMENTS ................... ....... 14-33 14-33 TS TS 4.1 Reactor Core Parameters Reactor Parameters ..........................
.......................... 14-34 14-34 TS 4.2 Reactor Reactor Control and Safety Safety Systems Systems .................
................. 14-36 14-36 TS 4.3 Coolant Coolant Systems System s ..................................
.................................. 14-38 14-38
- TS 4.4 Confinem ent .....................................
Confinement ..................................... 14-38 14-38 TS 4.5 Ventilation ............................... Ventilation Systems ............................... 14-38 14-38 TS 4.6 Emergency Electrical Emergency Electrical Power Systems ................ ................ 14-39 14-39 TS 4.7 Radiation Radiation Monitoring Monitoring Systems Systems and Effluents Effluents .................... 14-39 14-39 TS 4.7.1 Radiation Radiation Monitoring Monitoring Systems Systems Applicability Applicability ....... ....... 14-39 14-39 TS 4.7.2 Effluents ................................... Effluents ................................... 14-40 14-40 TS 4.8 Experim ents ..................................... Experiments ..................................... 14-40 14-40 TS 4.9 Facility-Specific Facility-Specific Surveillance ....................... Surveillance ....................... 14-40 14-40 TS 55 DESIGN TS FEATURES ........................................ DESIGN FEATURES ........................................ 14-41 TS 5.1 5.1 Site and Facility Description .........................
. . . . . . . . . . . . . . . . . . . . . .. 14-41 TS 5.2 Reactor Coolant ...........................
Coolant System ........................... 14-41 TS 5.3 Reactor Core Core and Fuel .............................
............................. 14-42 TS 5.4 Reactor Core .....................................
Reactor Core ..................................... 14-43 TS 5.5 ................................. Control Elements ................................. 14-43 TS 5.6 Fissionable Material Storage ....................... ....................... 14-43 TS TS 6 ADMINISTRATIVE CONTROLS 6 ADMINISTRATIVE CONTROLS .............................
. . . . . . . . . . . . . . . . . . . . . . . . . . . .. 14-45 14-45 TS 6.1 O rganization .....................................
Organization . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 14-45 14-45 TS 6.1.1 Structure Structure ...................................
................................... 14-45 14-45 TS 6.1.2 Responsibility ..............................
Responsibility. . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 14-45 14-45 TS TS 6.1.3 Staffi ng .................................... Staffing .................................... 14-47 14-47 TS 6.1.4 Selection and Training of Personnel Personnel .............
............. 14-47 14-47
- TS 6.2 Review Review and Audit .................................
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 14-47 14-47 UWNR Safety Analysis Report Rev. 2 TOC-8 TOC-8 Sept. 2008 2008
- TS TS 6.2.1 6.2.1 Composition Composition and Qualifications Qualifications.................14-48
................ 14-48 TS 6.2.2 TS 6.2.2 Charter Charter and Rules...........................
Rules ............................ 14-48 14-48 TS 6.2.3 TS 6.2.3 Review ........................... Function ............................ Review Function 14-48 14-48 TS 6.2.4 TS Audit Function............................. Audit Function . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 14-49 14-49 TS TS 6.3 6.3 Radiation Safety................................. Radiation Safety ..... , . . . . . . . . . . . . . . . . . . . . . . . . . . .. 14-49 14-49 TS 6.4 TS Procedures..................................... Procedures ...................................... 14-50 14-50 TS 6.5 TS 6.5 Experiment Experiment Review Review and Approval ................... Approval . . . . . . . . . . . . . . . . . .. 14-51 14-5 1 TS 6.6 TS 6.6 Required Actions Required Actions................................
................................. 14-52 14-52 TS 6.6.1 TS 6.6.1 Action to be Taken Taken in Case of Safety Safety Limit Violation 14-52 14-52 TS 6.6.2 TS 6.6.2 Action to be Taken Taken in the Event of an Occurrence Occurrence of the Type Identified 6.7.2(1)b., and 6.7.2(1)c Identified in 6.7.2(1)b., 6.7.2(1)c ........
........ 14-52 14-52 TS 6.7 TS 6.7 Reports .......................................
Reports ......................................... 14-53 14-53 TS 6.7.1 TS 6.7.1 Operating .......................... Operating Reports ........................... 14-53 14-53 TS 6.7.2 TS 6.7.2 Special Special Reports ............................ Reports ............................. 14-55 14-55 TS 6.8 TS 6.8 Records ....................................... Records ......................................... 14-56 14-56 TS 6.8.1 TS 6.8.1 Records Records to be Retained Retained for a Period of at least least Five Years or for the Life of the Component Component Involved Involved if Less than Five Years Years................14-56
................................. 14-56 TS 6.8.2 TS 6.8.2 Records to be Retained Retained for at Least Least One Cycle Cycle .......... 14-5614-56 TS 6.8.3 TS 6.8.3 Records to be Retained for the Lifetime of the Reactor Reactor Facility..............................................
Facility . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 14-56 14-56 TS 77 References References...............................................
................................................. 14-57
- TS 14-57 15 15 FINANCIAL FINANCIAL QUALIFICATIONS..................................
QUALIFICATIONS ................................... 15-1 15-1 15.1 Financial 15.1 Financial Ability to Construct a Non-Power Non-Power Reactor Reactor ...............
............... 15-1 15-1 15.2 Financial 15.2 Financial Ability to Operate a Non-Power Non-Power Reactor Reactor .................
................. 15-1 15-1 15.3 Financial 15.3 Financial Ability to Decommission Decommission the Facility Facility ....................
.................... 15-2 15-2 References................................................
15.4 References .................................................. 15-2 15-2 16 16 OTHER LICENSE CONSIDERATIONS ............................. LICENSE CONSIDERATIONS ............................ 16-1 16-1 16.1 Prior Use of Reactor 16.1 Components.............................. Reactor Components ............................... 16-1 16-1 16.2 Medical 16.2 Non-Power Reactors............................ Medical Use of Non-Power Reactors ............................. 16-1 16-1 Appendix A Appendix A Calculation Methods Calculation Methods for Atmospheric Atmospheric Release Radioactivity ....... Release of Radioactivity ...... A-1 A-I A. A. Models Used for Calculations Calculations in Sections Sections 11.1.1.1.2 11.1.1.1.2 and 13.1.1.5 13.1.1.5...... ..... A-A-II B. Sample Calculations Sample Calculations Supporting Supporting Section 11.1.1.1.2 ................. Section 11.1.1.1.2 . . . . . . . . . . . . . . . .. A-2 A-2 C. C. Calculations Supporting Calculations Supporting Section 13.1.1.4 (1), Section 13.1.1.4 (1), Whole Whole Body Body Exposure Exposure .. .. A-2 A-2 D. D. Calculations Supporting Calculations Supporting Section 13.1.1.4 13.1.1.4 (2), Dose to the Lungs ....... . . . . .. A-3 A-3 E. E. Calculations Supporting Section 13.1.1.5 Calculations 13.1.1.5 and Table Table 13.1 13.1 ............
. . . . . . . . . . .. A-3 A-3 References.....................................................
References ....................................................... A-4 A-4 Appendix B Appendix ....................................... Supporting Documents ......................................... B-I B-1 Task Order No.No.22 Under Under Master Task Agreement Agreement No. C96-175937 C96-175937 .......... .......... B-1 B-1
- UWNR Safety Safety Analysis Analysis Report Report Rev. 2 TOC-9 TC9Sp.20 TOC-9 Sept. 2008 2008
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- UWNR UWNR Safety Safety Analysis Analysis Report Report Rev. 2 TOC-10 Toe-IO Sept. 2008 2008
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- 1 1 THE FACILITY 1.1 1.1 Introduction Introduction This document document is prepared as part of an application for renewal of License R-74. R -74.
The University of Wisconsin has operatedoperated a teaching teaching and research research reactor, licensed licensed as R-74 under Docket 50-156 since 1961. The reactor since 1961. reactor supports teaching as a facility of the Engineering Physics Department, other departments departments of the university, and other educational institutions. Research Research use of the reactor supports the department, other University University of Wisconsin departments, numerous educational institutions, and some non-educational other educational non-educational groups. The reactor reactor is located on the campus of the university in a building located at 1513 University Avenue in Madison, Wisconsin. It currently operates operates at a 1000 1000 kW steady-state powerpower with pulsing capability capability to 1000 MW. The original Hazards Summary Report has been amended a number of times over the operating history, and the present present version of the Safety Analysis Report has been kept up to date by issuance replacement pages issuance of replacement pages at each each annual report submission. As a request for license renewal progressed, however, it was determined that the Safety Analysis Analysis Report should be replaced replaced with this completely new version structured structured in accordance accordance with the guidance guidance in NUREG 1537, "Guidelines "Guidelines for Preparing and Reviewing Reviewing Applications Applications for the Licensing of Non-Power Licensing Non-Power
- Reactors" Reactors" dated February 1996 1.2 Analysis 1996..
Summary and Conclusions Conclusions on Principal Safety Analysis of possible accident scenarios Considerations Safety Considerations scenarios is included in Chapter 13. 13. As a TRIGA-type TRIGA-type reactor, the primary safety features stem from the use of a fuel with a strong negative negative prompt temperature temperature coefficient coefficient of reactivity which limits excursions from reactivity insertions, thus preventing fuel damage damage from credible reactivity accidents. Ejection credible reactivity Ejection of the transient rod from the core when the core is operating at the power power level scram point will result in no fuel damage. Since experiments experiments are limited to the same reactivity worth as the transient rod, experiment experiment failure cannot cannot result in more severe transients. In addition, the operating operating power power level of 1000 1000 kW results in a decay heat potential in the fuel small enough enough that loss of reactor reactor coolant does not result in fuel damage or release of fission products. In extreme extreme accident conditions in which operation is taking place place with already already damaged damaged fuel, releases releases to the public are shown shown to be nominal except except for a combination of loss of pool water and loss of the ventilation system concurrent with the fuel damage. damage. Even with these extremely extremely conservative assumptions, conservative assumptions, analysis of this accident accident shows the event will result in exposure to the public that would be classified classified as an Alert. More realistic assumptions assumptions used in the accident
- UWNR UWNR Safety Analysis Report Rev. 2 1-1 1-1 2008 Sept. 2008
analysis indicates that the maximum hypothetical accident radiation radiation exposure within those allowed by 1.3 1.3 General Description of the Facility accident would result in emissions by 10 CFR Part 20. emissions and
- The University of Wisconsin Nuclear Reactor is located in the Mechanical Mechanical Engineering Engineering Building on the Madison campus of the University University within the city city of Madison, WI. The building also contains classrooms, classrooms, laboratories, laboratories, shops, and staff offices for the departments departments of Mechanical Engineering, Industrial Engineering, Engineering, and Engineering Physics departments.
Figure 1-11-1 is a pictorial view of the reactor. The reactor reactor is a heterogeneous pool-type pool-type nuclear reactor currently fueled with TRIGA-FLIP TRIGA-FLIP fuel modified to adapt to 4-element bundle assemblies. The coolant is light water which circulates through the core by natural convection. The core is reflected by water and graphite. Maximum steady-state steady-state power power level is 1000 KW. A 7 by 9 grid, surrounded by a core box, positions fuel, reflector, and control elements. Three shim-safety blades, blad.es, a transient transient control rod, and a regulating blade control core reactivity. The transient control rod is guided by a tube replacing a fuel element in a central fuel bundle, while the control blades blades move vertically vertically in two shrouds extending the length of the core. The grid box box and control element element drive mechanisms mechanisms are supported by a suspension suspension frame from the reactor reactor bridge. I I* Cold, clean core excess reactivity is about 4.3 % Ak/k. capability capability provide a shut-down margin of about 4.2 % Llklk. Proposed technical specifications Proposed technical Llklk. Control elements Ak/k. specifications for the facility are included elements having a scram included in this report as Chapter 14. 14.
*
SUMMARY
OF REACTOR
SUMMARY
REACTOR DATA Responsible Responsible Organization Organization The University University of Wisconsin Wisconsin Location Madison, Wisconsin Purpose Teaching Teaching and Research Research Fuel Type TRIGA and TRIGA-FLIP TRIGA-FLIP Hydride in Hydride in 44 element element clusters clusters Number of elements elements in standard 1000 KW Core UWNR U vv N R Safety S aiLULY Analysis Ana ys s Report Repoit Rev. ev. 2 I-_/ 1-2 Sept. /_VV6 2008
*
- Control Safety elements Safety elements Three vertical vertical blades Regulating-servo element Regulating-servo element One vertical blade Transient Transient control One rod rod Experimental Facilities Experimental Facilities Thermal column column One, 40-inch square graphite, 4)th =
= 2 x 10 1088 nv Beam Ports Four, 6-inch diameter diameter 4ý th
th== 1I -- 33 x 10101010 nv at shield side of shutter; about 88 x 10l 1011 at core end of port Pneumatic Pneumatic Tube One, 2-inch (sample(sample size 1/4 inch diameter by 11 114 5-1/2 5-112 inches long),
4)
th = = 4 x 10 12 nv.
1012 Hydraulic in-pool irradiation facilities Presently three, 2.5 inch (sample size up to 1 7/8 Presently 7/8
- diameter by 4 inches long) diameter
- 4) th = 8 x 10
101212 - 2.4 Xx 10\3 1013 nv, depending upon upon location location Thermal neutron fluxes for isotope production Thermal production include the above, plus large irradiation irradiation spaces outside the core with thermal neutron fluxes of around 1.3 x 1013 nv.
10\3 nv .
- UWNR UWNR Safety Analysis Report Rev. 2 1-3 1-3 2008 Sept. 2008
*
- 1-1 UWNR Figure 1-1 UWNR Open Pool Reactor Reactor Analysis Report Rev. 0 UWNR Safety Analysis 1-4 April 2000
*
- Dimensions Dimensions Pool 88 x 12 x 27-112 27-1/2 ft. deep Standard 1000 KW core 15 x 17 x 15 inches inches high high Grid box 9 x 77 array of 3-inch 3-inch modules Control blades 10-1/2 10-112 inches wide Fuel Element Element Diameter Diameter Length Length
--
Nuclear Fueled length Characteristics Nuclear Characteristics
-
- 1 MW Steady state:
Maximum Maximum thermal neutron flux 10"13 nv 3.2 x 10 Maximum Maximum fast neutron flux 3.0 x 1013 nv 1000 MW Pulse Maximum Maximum thermal 16 nv neutron flux 3.2 x 10 16 Maximum fast neutron flux 3.0 16 nv 3.0 x 10 16 Prompt temperature coefficient temperature coefficient 4 of reactivity -1.26 10-4 LiKlc
-1.26 x 10- AK/°C coefficient of reactivity Void coefficient 10-44 AK/%
-.2 x 10- LiKl% void void
- UWNR Safety Analysis Report Rev. 0a Analysis Report 1-5 1-5 April 2000
Prompt neutron lifetime Effective delayed Effective delayed neutron fraction 42 Ilsec p.sec TRIGA Standard 24 Ilsec pisec FLIP 0.007 Standard fuel,
- 1.4 Shared Shared Facilities and Equipment Equipment The Reactor Laboratory Laboratory and supporting laboratories are an integral part of the Mechanical Mechanical Engineering Engineering Building, and thus shareshare walls, water supplies, sewage, and main electrical distribution with the remainder of the building. Heating Ventilation & & Air Conditioning (HVAC)
(HV AC) systems are dedicated to non-shared non-shared use except for those HV HVAC AC systems in office spaces. The restricted restricted area contains contains only re~ctor-related reactor-related activities. 1.5 Comparison With Similar Facilities The best indication of reactor reactor characteristics characteristics is the performance performance of the facility itself, itself, which has been in routine operation operation with the present operational core since August, 1979. The reactor reactor at Washington State University is very similar to UWNR, having also originally been been a General General Electric Open Pool Reactor which was converted converted to TRIGA TRIGA fuel, and eventually eventually converted to FLIP fuel. The reactor at Texas partially converted Texas A & & M University is also a converted converted core,
- though the original reactor reactor was not built by GE. The pool size and experimental experimental facility configuration configuration differs on the three reactors, but basic reactor reactor behavior and accident accident analysis analysis are quite similar. In addition, the nuclear characteristics characteristics of UWNR ofUWNR are quite similar similar the TRIGA to Mark III prototype and other FLIPFLIP fueled reactors. In chapter 4 of this report the similarity between between the UWNR and the prototype is detailed.
1.6 1.6 Summary of Operations Operations Present plans and previous usage involve use of the reactor performance of the following reactor in performance experiments: experiments: 1.
- 1. Reactor Start-up and Operation;
- 2. Radiation Survey of the Reactor and Surroundings;
- 3. Control and Regulating Blade Calibration;
- 4. Measurement Measurement of Reactor Power and Calibration of Reactor Reactor Instruments;
- 5. Measurement Measurement of Shutdown Power Level; UWNR U W iNtý_ Safety Analysis tWPOR 3MCLY I'MMYSIS Report Rev.
MeV- 2
/- 1-0 1-6 Sept. 2008 2_008
*
- 6.
- 6. Measurement Measurement of Reactor Period,
- 7. Measurement of Temperature Measurement Temperature Coefficient Coefficient of Reactivity;
- 8. Measurement of Void Coefficient Measurement Coefficient of Reactivity; Reactivity;
- 9. Experiments Involving the Danger Coefficient Method; Experiments
- 10. Experiments Experiments to Measure the Disadvantage Disadvantage Factor; 11.
- 11. Studies of Reactor Buckling and Delta KlK.
Reactor Buckling K/K. 12.
- 12. Critical Mass Experiments; Experiments; 13.
- 13. Measurement Measurement of Thermal Neutron Cross Sections; Sections; 14.
- 14. Delayed Neutron Emission; 15.
- 15. Activation Analysis; 16.
- 16. Experiments Utilizing Pile Oscillator Experiments Oscillator Techniques;
- 17..
17 18. 19. Flux Distributions in Reactor and Effect of Absorbers Shielding Experiments; Experiments Experiments; Experiments on the Production Absorbers on Flux Patterns; Production of Radioisotopes; Radioisotopes; Patterns; 20.
*20. Neutron Diffractometer Diffractometer Measurements.
Measurements. 21.
- 21. Neutron Neutron Radiography The above represents represents the experiments planned at present, but it is anticipated that further experi-ments (both for training and research) will be added.
1.7 1.7 Compliance Compliance With the Nuclear Waste Policy Act of 1982 1982 In accordance accordance with a letter from the U. S. Department Department of Energy (R. L. Morgan) to the U. S. Nuclear Regulatory Regulatory Commission (H. (R. Denton) dated May 3, 3, 1983, 1983, it has been determined that all universities operating non-power universities operating non-power reactors have entered entered into a contract with DOE that provides that DOE retain title to the fuel and DOE is obligated to take the spent fuel and/or high level waste for storage or reprocessing. Because the University of Wisconsin Wisconsin has entered entered into such a contract contract with DOE, the applicable applicable requirements requirements of the Nuclear Waste Waste Policy Act of 1982 have
- UWNR UWNR Safety Analysis Report Rev. 2 1-7 1-7 Sept. 2008
satisfied by been satisfied Task Order under Master 15 60) is included 76ERO 1560) 1.8 1.8 University of Wisconsin by the University Master Task included in Appendix Facility Modifications Facility Wisconsin Nuclear Agreement Agreement Appendix B. Modifications and and History History Reactor Facility. A Nuclear Reactor (which is the successor successor to A copy contract copy of the current DE-ACO7-DE-AC07-current
- Construction permit CPRR-55, CPRR-55, authorizing construction construction of the University of Wisconsin Wisconsin Nuclear Nuclear (UWNR), was issued on June Reactor (UWNR), th 1960. License June 77 ",, 1960. License R-74 was issued November November 223,d, 1960.
3 "d, 1960. The expiration expiration date of the license was set at 40 years after after the issuance issuance of the construction construction permit permit (June 66 ', th
, 2000). The UniversityUniversity of Wisconsin Nuclear Nuclear Reactor Reactor achieved criticality on achieved initial criticality March 2 61h, 1961 March 26 t
\ 1961 as a 10 10 KW teaching teaching and research research reactor. After a license amendment dated license amendment October October 22 nd 1964, the power level was increased 2 2 nd,, 1964, increased to 250 250 KW on December December 77 t", \ 1964, 1964, using the original original flat-plate flat-plate aluminum clad fuel. Operations Operations withwith the original core ended ended October 13 13t,th ,
- 1967, 1967, after 2268.5 2268.5 critical 105.65 megawatt critical hours and 105.65 megawatt hours of core core exposure.
A A cooling system was installed installed and the reactor was converted converted to a 1000 1000 KW, TRIGA TRlGA reactor reactor with pulsing pulsing capability 1967. Construction capability in 1967. Construction permnit CPRR-97 authorizing permit CPRR-97 authorizing the changeschanges was issued on June June 7, 7, 1967. 1967. Amendment Amendment No. No.88 to the operating operating license for the conversionconversion was issued issued November 13, on November 13, 1967. 1967. Initial criticality with the TRIGA Initial criticality TRlGA core occurredoccurred on November November 14, 14, 1967. 1967. After After over 3,000 3,000 megawatt megawatt hours of operation operation with the TRlGA TRIGA core a partial refueling refueling was necessary. necessary. FLIP fuel was available available to afford significantly significantly improved improved core lifetimes, so a new Safety Analysis Report was submitted submitted in April, 1973 1973 describing describing facility characteristics characteristics and safeguards using standard safeguards standard fuel, FLIP fuel, and mixturesmixtures of the two two fuel types in defined defined
- compositions. License compositions. License amendment amendment No. No. 10 10 was was issued issued in in response.
response. The initial partial The initial partial refueling refueling to to a mixed core with 99 Standard Standard fuel bundles bundles replaced replaced with FLIP (Fuel Life Improvement Improvement Program)Program) fuel was initially critical in March initially critical March 1974. 1974. Additional Additional fuel replacements replacements in January January 19781978 and August 19791979 resulted resulted in the present present operating core, consisting entirely operating core, entirely of FLIP fuel. The total fuel exposure exposure since converting converting to TRlGA TRIGA fuel is over 20,000 megawatt hours, or 833 20,000 megawatt 833 megawatt megawatt days. days. A number of other license A license amendments amendments were issued during the term of the license, license, involving involving inclusion of security,security, training, and emergency emergency plans. Several other other amendments changed changed the amount and types of Special Special Nuclear Material used in connection Nuclear Material connection with the the license. None None of these changes had any effect effect on the operating operating characteristics characteristics of the reactor, reactor, andand are therefore therefore not detailed here. The most recent amendment amendment was was No. 16, 16, dated August August 30, 2006. 30,2006. A new Safety Analysis A Analysis Report was submitted submitted in April 2000 as part of a license extension extension application. application. Two Two changes changes were were included included in inthe the new new SAR: SAR: elimination elimination of of the the reactor reactor trip trip on on short short elimination of the electronic period and elimination electronic scram capability capability of the safety amplifier. These These changes changes were approved approved by by our Reactor Safety Committee Committee basedbased on a 10 10 CFR 50.5950.59 analysis. analysis. Neither Neither of of these items had been required by Technical Specifications, been required by Technical Specifications, since they did not provide protection not provide protection for a pulsing TRlGA Sections 7.2.3 TRIGA reactor. Sections 7.2.3 and 7.4 describe instrumentation as it has been describe the instrumentation changed. changed. UWNR UWNR Safety Analysis Report Report Rev. 2 1-8 18Sp.20 1-8 2008 Sept. 2008
*
- A new cooling system was installed in 2003. Revision 1 of the SAR was issued in August August 2004 2004 mainly to reflect the new cooling system, which is described in Chapter 5.
From 2004 through 2007 the Mechanical Mechanical Engineering surrounding the reactor laboratory Engineering building surrounding under went construction. The center wing was completely completely demolished demolished and rebuilt, and the east, west, and north wings, as well as the reactor reactor auxiliary support spaces, were substantially substantially renovated. Some reactor systems werewere also renovated, but the reactor itself remained unchanged. Revision 2 of the SAR was issued in September September 2008 to address the relevant changes. The new new Chapter 2. A new ventilation system was installed, described building floor plans are shown in Chapter primarily in Chapter 9 (the calculations in Chapter 11, Chapter 13, 11, Chapter 13, and Appendix A were were also affected). components were installed as part of a console affected). Several new reactor control console components instrumentation instrumentation upgrade, and Chapter 7 was updated to reflect reflect these changes. A new pneumatic tube sample transfer system (which mates with the existing irradiation irradiation facility) was installed installed and described in Chapter 10. These changes is described changes were approved by our Reactor Safety Safety Committee based based on a 10 CFR 50.59 analysis . analysis.
- UWNR UWNR Safety Analysis Report Report Rev. 2 1-9 1-9 2008 Sept. 2008
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- UWNR Safety UWNR Safety Analysis Analysis Report Report Rev. 2 1-10 1-10 2008 Sept. 2008
- 0
- 2 2.1 2.1.1 SITE CHARACTERISTICS SITE CHARACTERISTICS Geography and Demography Geography Demography Site Location and Description Description 2.1.1.1 Specification and Location Location The University of Wisconsin Nuclear ReactorReactor Laboratory is located located within the Mechanical Mechanical Engineering Engineering Building at 1513 University Avenue on the University of Wisconsin-Madison Wisconsin-Madison (UW) campus. The reactor is located at States Geological On the United States Survey (USGS) Madison Madison West, WIS 15' 15' Quadrangle topographical map, the Universal Transverse Mercator Mercator Coordinates are; .
The UW campus is surrounded by the city of Madison Madison in Dane county, Wisconsin. Figure 2-1 2-1 shows the location of Dane county within Wisconsin. Madison, a city of approximately approximately 208,000 residents (2000 Census statistics), is built around two lakes in the center of Dane county, Figure 2-2. Lake Mendota (15 (15 square miles) lies northwest of Lake Monona (5 square miles) and the two lakes are only 2/3 of a mile apart at one point. The UW campus is set on this narrow narrow neck ofof land between the two lakes, known as the isthmus, and on the southern southern shore shore of Lake Mendota. The Mechanical Engineering Mechanical Engineering Building is near the southwestern southwestern border of the UW campus, where the nearest non UW owned owned property is 425 ft (130 (130 m) from the reactor reactor site (approximately (approximately 80m80m
- from the Mechanical Engineering Building west wall). The reactor Mechanical Engineering the shore of Lake Mendota.
2.1.1.2 Boundary and Zone Area Area Maps reactor is 2300 ft (700 m) south of of A map of the City of Madison Madison detailing the general topography and the surrounding surrounding urban and rural zones up to a distance distance of 88 km is reproduced reproduced in Figure 2-3. The UW campus, which is located on the southern southern shore of Lake Lake Mendota, is shown in Figure 2-4. The Mechanical Engineering Building is located on the engineering campus Engineering Building campus which is the southwest corner comer of the UW campus as shown in Figure 2-5. The operations boundary is'defined is'defined as the Reactor Laboratory, Room 1215, of the Mechanical Room 1215, Engineering Building. The site boundary boundary is defined as the center and east wings of the Mechanical Mechanical Engineering Engineering Building, but not including the north or west wings, plus the portion of of Engineering Engineering Drive (formally Johnson Johnson Drive) south. south of the designated designated areas of the building. Figure 2-6, Figure 2-6, Figure 2-7, Figure 2-8, Figure 2-9, Figure 2-10, and Figure 2-11 2-1,1 depict the floor plans of the Mechanical Engineering Building's basement, first floor, second second floor, third floor, fourth floor, and fifth floor respectively. Figure 2-12 and Figure 2-13 2-13 depict cross sections of of the building through the core centerline centerline looking north and east, respectively. The emergency emergency preparedness preparedness zone is entirely entirely within the operations boundary, as defined above. above .
- UWNR Safety Analysis Report Rev. 2 2-1 Sept. 2008 2008
- Figure 2-1 Wisconsin Map
'~~~~~~ vruA 3 caI N &
1.*n 0 2.1.2 Population Distribution Figure Figure 2-2 Distribution Dane County, Wisconsin
- uniform density model and the 2000 Census11 are shown Population distributions estimated by the uniform residential, and the central business district 2-1. The area around the campus is mature residential, in Table 2-1.
is only a short distance short distance away. Population is therefore quite stable in the immediate surrounding surrounding areas of the UW campus. The 8 kilometer kilometer radius sparsely-settled regions to radius includes much more sparsely-settled the north of Lake Mendota, and population population in this region is likely to increase markedly in the future. The nearest permanent approximately 150 m west of the reactor permanent residence is approximately reactor site. Transient population Transient reactor include students popUlation around the reactor students present classrooms and offices in the present in classrooms Mechanical Engineering Building during Mechanical Engineering Building during the months of September September through May as well as Mayas spectators attending sporting events at Camp spectators attending Camp Randall StadiumStadium which approximately 250 m due which is approximately south south of the reactor. The maximum number The maximum t:J.umber of students students present present in the east and center the east center wings wings of of the Mechanical Engineering Building Mechanical Engineering Building at any anyone estimated at 1500. one time is estimated 1500. The maximum maximum number number of spectators spectators contained contained by Camp Camp Randall Stadium Stadium is 80,321. 80,321. . UWNR UWNR Safety. Safety Analysis Analysis Report Report Rev. 2 2-2 2-2 Sept. 2008 2008
*
- Figure 2-3 City ooff Madison, Wisconsin
- 2-3 Wisconsin UWNR Safety Analysis Report Rev. 2 2-3 2-3 Sept. 2008 2008
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Figure 2-4 2-4 University of Wisconsin Campus UWNR Safety A Analysis nalysis Report Rev. 2 2-4 2008 Sept. 2008
*
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- UWN UWNR R Safety Safety Analysis Report Rev. 2 2-55 2* 2008 Sept. 2008
*
- Figure 2-6 2-6 Mechanical Engineering-Mechanical Engineering Basement Basement UWNR Safety Analysis Report Rev. 2 UWNR 2-6 2-6 2008 Sept. 2008
*
- Figure Figure2-7 Mechanical Engineering
- 2-7 Mechanical Engineering --First FirstFloor Floor UWNR UWNRSafety SafetyAnalysis AnalysisReport Report Rev.
Rev. 22 2-7 2-7 Sept. Sept. 2008 2008
*
- Figure 2-8 Figure 2-8 Mechanical Engineering - Second Floor Mechanical Floor UWNR Safety Analysis Report Rev. 2 2-8 2-8 Sept. 2008 2008
*
- 2-9 Figure 2-9 Mechanical Mechanical Engineering Engineering - Third Floor
- UWNR UWNR Safety Analysis Analysis Report Rev. 2 2-9 2-9 Sept. 2008
I I I I I
- I I
I I I I I I I I I I I I I I I I I I I I
- I I
I I t I I t I I t I I Figure 2-10 Mechanical Engineering - Fourth Floor t Figure 2-10 Mechanical Engineering - Fourth Floor UWNR UWNR Safety SafetyAnalysis Analysis Report ReportRev. Rev. 22 2-10 2-10 Sept. 2008 Sept. 2008
*
- I I
I I I I I I I I
- Figure Figure2-11 2-11 Mechanical Mechanical Engineering Engineering -- Fifth FifthFloor Floor
- UWNR UWNR Safety Safety Analysis Analysis Report ReportRev.
Rev. 22 2-11 2-11 Sept. Sept. 2008 2008
- 0
- Mechanical Engineering Cross Section through Core Figure Figure2-12 2-12 Mechanical Engineering Cross Section through Core
- Centerline, looking North Centerline, looking North UWNR UWNR Safety Safety Analysis Analysis Report ReportRev.
Rev. 22 2-12 2-12 Sept. 2008 Sept. 2008
- Figure Figure2-13 2-13 Mechanical Mechanical Engineering EngineeringCross Cross Section Sectionthrough through Core CoreCenterline, Centerline, looking lookingEast East
- UWNR UWNRSafety SafetyAnalysis AnalysisReport ReportRev.
Rev.22 2-13 2-13 Sept. Sept.2008 2008
Distance TABLE 2-1 Population Distribution Population Distance from Facility(kilometers) Facility(kilometers) Population 1 Estimated 2000 Population'
- 0 1 12,595 2 37,814 37,814 4 71,780 71,780 6 118,919 118,919 88 192,575 192,575 2.2 Nearby Industrial, Transportation, Transportation, and Military Military Facilities 2.2.1 2.2.1 Locations Locations and Routes The UW campus campus is surrounded, mainly, by a residential district to the south and west, while to
- the east is primarily a commercial commercial business business district with city and state government government office office buildings. No industrial industrial facilities are in the vicinity of the reactor.
A railroad spur of the Wisconsin Wisconsin & & Southern Southern Railroad Railroad Company runs through campus and is 100m to the Reactor Reactor Laboratory at its closest approach. A rail car holding holding yard which is a part of of this spur is approximately approximately 300 m northwest. The primary commodity commodity transported transported over this spur 2. or resident resident at the rail car holding yard is coal coaf. The reactor reactor is located approximately approximately 4.5 km north of a bypass highway, known as the Beltline,Beltline, for US highways 12, 14, 18 and 151. 151. Interstates 90 and 94 are located approximately approximately 10 km to the east and northeast northeast of the reactor site. There are no military facilities in the Madison Madison area with the exception exception of the Wisconsin Wisconsin Air National National Guard which is located on a military ramp of the Dane County Regional Regional Airport. While the Wisconsin Air National National Guard flies approximately approximately 4000 missions annually, the flight pattern of these missions missions typically are north of Lake Mendota and the City of Madison. At no time do ammunition33.* More information about the Dane County any of these mission flights carry live ammunition Regional Airport Airport facility is reported in section 2.2.2 Air Traffic. UWNR UWNR Safety Analysis Report Report Rev. 2 2-14 2-14 Sept. 2008 2008
*
- 2.2.2 Air Traffic The Dane County Regional Airport is approximately approximately 8 km to the north east of the reactor site.
This is the only commercial commercial airport near the reactor. While there are three smaller air fields within 16 km from the reactor reactor in the communities surrounding Madison, these air fields are for general aviation only. The Dane County County Regional Regional Airport has 3 runways with the following outbound headings; 0 360°(north)/1800, 360 320'/140', and 2100/300. (north)1180°, 320°1140°, 210°/30°. None of these headings have trajectories that take commercial commercial traffic directly directly over the reactor reactor immediately before. before arrival or after departure. The airport is serviced by several commercial commercial express carriers and is utilized by the Wisconsin Wisconsin Air Air National 3 3 National Guard as well as general aviation aircraft *. Of the 115,613 events (arrival and departure are counted as two separate events) in 2006, 59% were classified 2006,59% classified as general aviation, 34% carriers/taxi, and 7% commercial carriers/taxi, 7% military. Due to the infrequent arrival of commercial traffic, the air traffic control tower tower does not place inbound inbound traffic in holding patterns around the city of of Madison4.44* Madison 2.2.3 Analysis of Potential Accidents at Facilities Facilities There are no industrial, transportation transportation or military facilities within the vicinity of the reactor site that have the potential for accidents with consequences consequences significant to impact the Reactor Reactor
- Laboratory. While a railroad spur passes within 100 m of the reactor facility, this spur transports non hazardous hazardous cargo and other major ground transportation transportation routes are located at great distances distances from the Reactor Reactor Laboratory. Due to the frequency frequency and flight paths of commercial commercial and military air traffic the probability occurrence of an accident probability of occurrence accident is considered considered extremely low.
2.3 Meteorology Meteorology 2.3.1 General and Local Climate continental climate of interior North America with a large annual Madison has the typical continental temperature temperature range and with frequent short period temperature changes. The range of extreme temperatures is from about 110 to -40 OF. 'F. Winter temperatures temperatures (December - February) average near 20 OF'F and the summer summer (June - August) average average temperature temperature is in the upper 60s. Daily temperatures average average below 32 OF 'F about 120 days and above 40 'F OF for about 210 days of the year. Madison Madison lies in the path of the frequent cyclones cyclones and anticyclones anticyclones which move eastward over this area during fall, winter and spring. In summer, the cyclones have diminished intensity and tend to pass farther north. The most frequent air masses are of polar origin. Occasional outbreaks of arctic air affect this area during the winter months. Although northward moving tropical tropical air masses contribute considerable cloudiness contribute considerable cloudiness and precipitation, the true Gulf air mass
- UWNR Safety Analysis Report Rev. 2 2-15 2-15 2008 Sept. 2008
does not reach this area in winter, and only occasionally with only occasional occasionally at other seasons. Summers occasional periods of extreme heat or high humidity. There are no dry and wet seasons, but about 60 percent percent of the annual months of May through September. Cold season precipitation Summers are pleasant, precipitation falls in the five annual precipitation precipitation is lighter, but lasts longer. During
- July, August, and September rainfall is mostly from thunderstorms and tends to be erratic and variable. Average occurrence occurrence of thunderstorms thunderstorms is just under 7 days per month during this period.
Tornadoes Tornadoes are infrequent. Dane County has about one tornado in every three to five years. covered with 1 inch or more of snow about 60 percent The ground is covered percent of the time from about December December 10 to near February February 25 in an average average winter. The soil is usually frozen from the first of of inches.",5 with an average frost penetration of 25 to 30 December through most of March with an average frost penetration of25 to 30 inches."s . December 2.3.2 Site Meteorology Meteorology The summary of meteorological meteorological conditions for Madison is based on the records obtainedobtained from the International Meteorological Climate Summary6 International Station Meteorological Summary6 jointly jointly produced by the National Oceanic and Atmospheric Atmospheric Administration Administration (NOAA), (NOAA), the United States Air Force (USAF) and specifically compiled United States Navy. The data specifically compiled for Madison was obtained obtained from the National Weather Weather Service and unless specifically noted, is for the period of record record from 1948 to 1995.
- The Reactor Laboratory does not have a continuing onsite meteorological meteorological data measurements measurements program. All future meteorological data will be obtained obtained from the National Weather Service station in Madison.
2.3.2.1 Temperature Temperature The monthly average and daily average extreme temperatures temperatures for the Madison area are shown in Table 2-2. The record extreme temperatures temperatures in the Madison Madison area, as reported by the National Weather Service, have ranged from a low of-37 Weather Service, of -37 'F OF in January 1951 to a high of 107 'F OF in July 1936. 1936. 2.3.2.2 2.3 .2.2 Precipitation Precipitation The Madison area normally receives an annualannual average average of 31.6 inches of precipitation. precipitation. The monthly average precipitation precipitation data is reported in Table 2-3. The record maximum maximum level of of precipitation precipitation to fall in Madison Madison in one year, is reported by the National Weather Weather Service to be 1881. The greatest 24-hour rain fall total for Madison 52.9 inches in 1881. Madison was 5.25 inches on July 15- 15-16, 1950. The 48-hour, 100-yr. return period rainfall for south central Wisconsin is estimated to be 7.82 inches77 .* The annual average snow fall for Dane County is 43.7 inches. The monthly average snow fall data is reported in Table 2-4. The record maximum snow to fall during the winter season is UWNR Safety Safety Analysis Analysis Report Report Rev. 2 2-16 2-16 2008 Sept. 2008
*
- reported by the National Weather Service as 76.1 inches during during the winter of 1978-79. The 24 24 hours state record for heaviest heaviest snow occurred occurred December 27-28, 1904 in Neillsville, Neillsville, Wisconsin in Wisconsin which 26 inches inches of snow snow fell.
TABLE 2-22-2 Average Temperatures Temperatures for the Madison Area Average Monthly Monthly Average Average Daily Average Daily Temperature Temperature Maximum Temperature Maximum Temperature Minimum Minimum Temperature (OF) COF) (OF) COF) (OF) COF) January 16.8 25.7 8.0 8.0 February February 21.3 30.6 12.0 12.0 March 32.3 41.9 22.7 April 46.1 57.6 34.7 34.7 May May 57.5 70.0 44.9 June 67.0 79.4 54.5
- July August September September 71.4 69.2 60.4 83.3 81.0 72.3 59.5 57.4 57.4 48.6 48.6 October 49.5 60.9 38.2 38.2 November November 35.3 44.0 26.5 December 22.6 30.7 14.4 14.4 Year 45.8 56.5 35.1 35.1
- UWNR Safety Analysis Report Rev. 2 2-17 2-17 2008 Sept. 2008
TABLE 2-3 Monthly Precipitation Data for the Madison Area Average Monthly Total Precipitation Precipitation Average Average Monthly Maximum Precipitation Precipitation
- 0 (inches)
(inches) (inches) (inches) January 1.14 2.45 February 1.14 2.77 March 2.17 5.04 5.04 April 3.02 7.11 May 3.14 6.26 6.26 June 3.83 9.95 July 4.05 10.93 August 3.90 9.49 September September 3.12 9.22 9.22 October November November December 2.29 2.15 1.66 5.63 5.13 5.l3 4.09
- UWNR Safety Analysis Report Report Rev. 2 2-18 2-18 Sept. 2008
*
- TABLE 2-4 2-4 Monthly Snowfall Snowfall Data for the Madison Madison Area Area Monthly Average Average Monthly Maximum Year Total Snowfall Snowfall (inches) (inches)
January 10.3 31.8 1929 1929 February 7.7 37.0 1994 1994 March March 8.5 28.4 1923 1923 April 2.3 17.4 17.4 1973 May 0.1 5.0 5.0 1935 1935 June 0 July 0 August 0 September Trace
- October October November November December December 0.2 3.8 10.9 5.0 18.3 32.8 1917 1917 1985 1985 1987 1987 2.3.2.3 Winds The average annual wind speed speed in the Madison area is 9.8 mph. The prevailing prevailing winds during the months of November November through March March are from the west-northwest west-northwest direction, the remaining months of April though October the prevailing prevailing winds are from the south'.
souths. Table 2-5 reports the frequency of surface wind direction direction versus wind speed. The record area record gust in the Madison area occurred in June 1975 when wind gusts were reported at 83 mph from the west.
- Safety Analysis Report Rev. 2 UWNR Safety 2-19 2-19 2008 Sept. 2008
Direction Frequency Frequency of Surface TABLE 2.:5 Wind 2-5 Direction Speed (knots) versus Wind Speed 1 --33 4-6 7-10 11-16 17-21 22-27 28-33 34-40 41-47 >=48 Percent Speed Percent Wind
- Speed Speed (knots)
N 0.5 1.4 1.6 1.3 0.2 ** ** ** 00 00 4.7 8.2 8.2 NNE 0.3 0.7 1.1 0.8 0.2 ** ** 0 0 00 3.3 9.3 NE NE 0.6 1.1 1.4 1.1 0.2 ** ** 00 00 0 4.6 9 ENE 0.6 1.7 1.7 1.2 0.2 ** ** 00 00 0 5.5 5.5 8.2 8.2 E 0.5 0.51 1.1 1.4 0.9 0.1 ** ** ** 00 0 3.8 7.8 7.8 ESE 0.3 0.9 1.2 0.7 0.31 0.1 ** 0 00 0 0 3.2 8.5 SE 0.4 1.1 1.5 0.7 0.1 ** 0 00 ** 0 3.8 8.5 8.5 SSE 0.4 1.4 104 1.7 1 0.1 0.1 ** 0 00 0 0 4.9 8.4 804 S S 0.6 2.8 4.1 2.9 0.5 0.1 ** ** ** 0 10.4 lOA 8.8 8.8 SSW 0.3 1.1 2.4 2.3 0.4 0.1 ** ** 0 0 6.8 ' 10.1 10.1 SW 0.3 1.1 2.4 2 0.4 0.1 * * ** ** 0 6.5 10.3 10.3 WSW WSW 0.3 1.1 2 1.5 0.3 0.1 ** ** ** 0 5.5 10 W. 0.4 1.6 2.5 1:9 1.9 0.5 0.1 ** ** 0 00 6.7 9.5 WNW NW NNW VAR CLM 0.4 1.9 2.9 2.6 0.6 1.6 2.4 2.5 0.5 0 0 1.6 1.6 1.6 0 0 00 00 00 00 0.6 0.6 0.3 0 0 0.1 0.1 0.1 0 00 1
**
**
00 00
**
0
**
**
*
- 0 0
0
*
- 0 00 00 00 00 00 00 00 8.6 8.3 5.8 0
7.6 100 9.7 9.7 10.1 10.1 8.5 9 0 0
- 0 ALL 6.9 22.2 32.1 25.2 4.7 0.1 ** *
- 100 8.5
* == PERCENT <<.05 .05 2.4 Hydrology Madison is located just northeast of the "Driftless "Drifiless Area" of southwestern southwestem Wisconsin. The glacially shaped topography of Madison and the surrounding surrounding area in central Dane County is irregular, ranging from flat or gently rolling to hilly. The most prominent geomorphic features were glacially formed. Among these are Lakes Mendota, Monona, Waubesa, Kegonsa and 9.
Wingra. Glacial drift covers the entire area except for local areas of bedrock bedrock outcrop outcrop9. In the vicinity of the reactor, a glacial deposit exists which contains clay and large boulders. Although this deposit may be as much as 100 feet thick, it is probably less than twenty. The reason for the variation in thickness is that the bed-rock sandstone which underlies the deposit is very uneven. The bed-rock bed-rock consists of Cambrian sandstones which are 700 to 800 feet thick and which are permeable to water. UWNR Safety Analysis Report Rev. 2 2-20 Sept. 2008
*
- River is The Yahara River is the main drainage feature in the Madison area. The Yahara River gradient is through the lakes essentially flat where it flows through lakes in Madison. Water surface surface elevations are controlled mostly by damsdams at the outlets oflakes of lakes Mendota and Waubesa.
Waubesa. The ground water flow from the reactor site is from is generally toward the Lake Mendota -- Yahara River -- Lake Lake Monona system. Thus, the general flow is toward the east and south. Historical data obtained from the USGS Wisconsin Geological and Natural History SurveylO Survey1" indicated the annual average water table in the vicinity of the reactor is approximately 60 feet below the surface. Madison obtains its drinking water supply from several deep well aquifers drilled, typically several hundred feet, into the Cambrian sandstone described above. The location of these wells is shown on Figure 2-12, and they supply the University as well as the city. All of these wells are cased from ground level into the sandstone so as to keep out ground water from the glacial deposit. The closest well to the reactor is Madison City Well 27 located 2,000 feet southeast. Due to the large drainage capacity capacity of the Yahara river and the outlet dams of lakes Mendota and Wausbesa flooding is not a serious problem in most of Madison. The 100 year flood is estimated to increase the surface level of Lake Mendota approximately 3 feet causing the lake to over flow feet11 . its existing banks by about 30 feet!!
- UWNR UWNR Safety Safety Analysis Analysis Report Report Rev. 2 2-21 Sept. 2008 2008
*0
*
- Figure 2-12 Madison Well Water City of Madison Water Supply
,
UWNR Safety Analysis Report Rev. 2 2-22 2-22 Sept. 2008
- 2.5 2.5 Geology, Seismology, and Geotechnical Engineering 2.5.1 2.5.1 Regional Geology Regional The Midwestern Midwestern United States does not lie on, or anywhere near, a tectonic plate boundary.
region is in the middle of The region of the North American Plate, hundreds of miles from both the (NO (A NA eastern and western edges. The ILLINOIS Midwest, however, has a series of faults around the Mississippi Valley, Figure 2-13, the most MISSOURI active of which is the New Madrid Fault System. These faults were formed by the tearing open of the ancient continental crust almost 5 million years ago. This region is known as the New Madrid
- Seismic Zone which includes includes the states of Missouri, Missouri, Arkansas, Arkansas, Louisiana, Illinois, Indiana and parts of Kentucky Kentucky and Tennessee1 122 .* Wisconsin Tennessee Wisconsin is not associated with this region. The New Madrid Fault System, ***
located near New Madrid, ! TENNESSEE Missouri is greater greater than 500 500 **
**
miles from Madison, Madison, Wisconsin Wisconsin ** and poses little or no seismic hazard. MtSS1SSIPPI *:* ALABAMA
**
,/' **
**
**
Figure Figure 2-13 New Madrid Seismic Zone
- UWNR UWNR Safety Safety Analysis Analysis Report Report Rev. 2 2-23 2-23 Sept. 2008 2008
2.5.2 Site Geology As discussed in section 2.4 Hydrology, the deposits which rests on the Cambrian glacial deposits the local geology of the reactor site includes a layer Cambrian sandstone bedrock. The layer of the glacial deposit layer of approximately 16 feet is variable, however, soil bores at the reactor site indicate this layer is approximately 13 of
- 0 deep, Table 2_6 ,14,15. The Cambrian sandstone layer below the glacial drift is approximately 2-613'14,11.
700 to 800 feet thick. Below the sandstone is impermeable basement rock. There are no geological structures of consequence in the vicinity of the reactor site. TABLE 2-6 2-6 Depth of Glacial Deposit to Cambrian Sandstone Distance from Reactor Site Distance Depth of Glacial Deposit (Feet) (Feet) 0 16 16 150 150 16.5 16.5 175 175 4
- 200 14 14 250 250 13.25 13.25 300 9.5 9.5 350 17.5 17.5 400 15 15 450 9.5 9.5 2.5.3 Seismicity Seismicity As discussed discussed in section section 2.5.1, 2.5.1, Regional Regional Geology, Wisconsin Wisconsin is located in a geologically geologically stable region of the United States.States. The closest closest active fault is greater greater than 500 miles in the New Madrid Madrid Seismic Seismic Zone. While no seismic events have occurred occurred at the reactor reactor site, there there are records of records of earthquakes earthquakes occurring within 200 km of the reactor site. A review of the USGS earthquake earthquake database" 6 16 database for all earthquakes earthquakes of modified modified Mercalli Mercalli intensity intensity greater then then IV or or magnitude magnitude greater greater than 3.0 which is within within 200200 km km of the reactor reactor site resulted in the data reported reported in Table 2-7.
UWNR UWNR Safety Safety Analysis Analysis Report Report Rev. 2 2-24 Sept. 2008 2008
*
- TABLE TABLE 2-7 2-7 Events Within 200 km of Madison, Wisconsin.
History of Seismic Events Date Magnitude Magnitude Intensity Distance (Richter) (Modified Mercalli) (Modified Mercalli) (km) August 20, 1804 4.4 VI 178 178 May 27, 1881 4.6 VI 198 198 May 26,1909 26, 1909 5.1 VII 195 195 January 2, 1912 4.5 VI 190 190 November 12, 12, 1934 1934 4.0 VI 196 196 September September 15,15, 1972 4.5 VI 158 158 September 9, 1985 3.0 V 178 178 2.5.4 Maximum Earthquake Potential Maximum Earthquake
- earthquake potential and frequency is based on data from previous determination of earthquake The determination events..
previous events earthquakes in the vicinity Because there are few historic moderate to large earthquakes vicinity of the reactor, earthquake potential may be inferred from data supplied by analysis is difficult. The maximum earthquake by 7 the USGS for the 50 year peak ground acceleration estimate 17.. The estimated 50 year peak acceleration estimate" acceleration due to a seismic event in the vicinity at the reactor is less than 0.01 g. ground acceleration 2.5.5 Vibratory Ground Ground Motion Motion insufficient data from previous seismic events the vibratory ground motion can Due to insufficient can only be section 2.5.4 of less than 0.01 g. inferred from the peak ground acceleration data of section inferred 2.5.6 Surface Faulting Based distance to any known active faults and the stable site geology, surface faulting is Based on the distance vicinity of the reactor not considered to be a credible event in the vicinity reactor site. 2.5.7 Liquefaction Potential liquefaction Based on the distance to any known active faults and the stable site geology, the liquefaction potential is considered considered to be insignificant.
- UWNR Safety Analysis Report Rev. 2 2-25 2-25 Sept. 2008 2008
2.6 2.6 1. References References Prepared Prepared by the Applied Applied Population Population Laboratory, generated by Applied Population generated Population Laboratory UW-Madison/Extension. Source: Laboratory, UW-MadisonlExtension. Bureau; 2000 Census of Population and Housing, summary File 1 (SF1); Census Bureau; Source: U.S. (SF1); Laboratory using Census 2000 Summary File 1 (SF1) (SF1)
- Wisconsin CD-ROM.
- 2. Meighan, Meighan, Ben. Union and Southern Southern Railroad Company. Telephone Telephone Interview.
January 24, 2000.
- 3. of Operations/Public Safety, Dane County Regional Airport.
Lenss, Marty. Director ofOperations/Public Email Interview(lenss@msnairp011.com). Interview (lenss@msnairport.com). March 27, 2007. 2007.
- 4. Donovan, Greg. Telephone Telephone Interview. February 5, 5, 2000.
5.
- 5. International International Station Meteorological Summary. Version 4.0. US ISMCS Station Meteorological Climate Summary. Station Climatic Narrative. National Climatic National Climatic DataData Center, September 1996.
- 6. International Meteorological Climate Summary.
International Station Meteorological Summary. Version 4.0. US ISMCS StationStation Database. National Climatic Data Center, September September 1996. 7. 8. Huff, F.A., and Angel, J.R Knox, P.N. Wind Bulletin Bulletin 94. 1996. J.R.... Rainfall Climate Center (MCC) Research Rainfall Frequency Research Report Frequency Atlas of the Midwest. Midwestern Report 92-03. 1992. Wind Atlas of Wisconsin. Wisconsin Geological Geological and Natural Midwestern Natural History Survey.
*
- 9. Flood Insurance Insurance Study. Federal Federal Emergency Emergency Management Management Agency. Community Number 550083. March 5, 550083. 5, 1996.
- 10. Observation Network. Wisconsin Groundwater Observation Wisconsin Geology Geology and Natural Natural History Survey.
September 13, September 13, 1999. 1999. http://wi.water.usgs.gov/gw/ http://wi.water.usgs.gov/gw/ 11.
- 11. National National Flood Insurance Insurance Program. FloodFlood Insurance Insurance Rate Map Community Panel Number Number 550083 0010 D. March 8, 5500830010 8, 1974.
12.
- 12. Atkinson, William. The Next New Madrid Earthquake. Southern Illinois University Earthquake. Southern Press. Carbondale IL,IL,1989.
1989. 13.
- 13. Engineering Engineering Centers Facility Facility Soil Boring Log. Woodward-Clyde Woodward-Clyde International-America.
International-America. Waukesha, WI. July 9, 1998. 1998. UWNR Safety Analysis Report Report Rev. 2 2-26 2-26 Sept. 2008
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- 14..
14 Preliminary Preliminary Soils Exploration of Lot 17 Parking Structure. Soils && Engineering Engineering Services Inc. Madison, WI. August 4, 1992. 1992.
- 15. Subsurface Exploration Exploration of Proposed Lot 17 Parking Structure.
Structure. Terracon Consultants Inc. Naperville, Naperville, IL. August 18, 18, 1997.
- 16. National National Earthquake Earthquake Information Center. United States Geological Geological Survey.
January 18, 2000. http://wwwneic.cr. http://wwwneic.cr.usgs.gov/neis/epic/epic.html usgs.gov/neis/epic/epic.html
- 17. National National Seismic Hazards Mapping Project. United States Geological Geological Survey.
19, 1999. http://geohazards.cr.usgs.gov/eq November 19,1999. http://geohazards.cr.usgs.gov/eg
- UWNR Safety Analysis Analysis Report Rev. 2 2-27 2-27 Sept. 2008 2008
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- UWNR UWNR Safety Safety Analysis Report Report Rev. 2 2-28 2-28 2008 Sept. 2008
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- 33 3.1 3.1 DESIGN OF STRUCTURES, DESIGN Design Criteria STRUCTURES, SYSTEMS, When the University of Wisconsin Nuclear SYSTEMS, AND AND COMPONENTS COMPONENTS Nuclear Reactor was to be upgraded by increasing authorized authorized power to 1000 kW the principal designdesign criterion was to assure the facility could withstand loss of of pool water and any other credible credible accident with no hazard to the public, without reliance on on engineered engineered safety features. This criterion criterion was met by selecting stainless-steel-clad stainless-steel-clad TRIGA TRlGA fuel due to the well-documented well-documented characteristics type l . Details characteristics of this fuel type'. Details of the physical mechanisms ofthe characteristics that cause TRlGA and characteristics TRIGA and FLIP fuel to exhibit the prompt negative temperature coefficient responsible for the fuel characteristics characteristics are given in the reference reference and a number of of other documents and are not included here. DetailedDetailed analysis for this facility (Chapter of this (Chapter 13 ofthis report) agree with the conclusions in the reference. The design criteria that result in this negligible safety risk are the result of the fuel composition composition and cladding, not of specific features provided in the equipment and building that surrounds the reactor.
Engineering Building which houses the reactor The Mechanical Engineering laboratory was completed reactor laboratory completed in 1930. Extensive remodeling of the north, east, and west wings, and construction of ofaa new center wing was completed in 2007. Extensive remodeling of the room that became the Reactor Reactor Laboratory Laboratory took place in 1960 when the reactor was installed. The original reactor installation installation used fuel and components components manufactured manufactured by General Electric, and the specifications specifications to which which
- structures were built were those stated by General Electric.
specifications and are detailed in section 9.1. Electric. Specific design criteria were not stated. The original ventilation system was designed in 1962 and installed in 1963. specifications stated only the desired desired flow rates and stack height. The current ventilation system was installed in 2006. The design specifications included included the desired flow rates and stack height, 9.1. The original cooling system was designed in 1966 and installed not 1963. The design installed 1967. The only design specification in 1967. specification was the heat removal removal rate required. The current cooling system was installed in 2002. The design specifications specifications included the heat removal rate required, and are detailed in chapter detailed chapter 5. 5. During the 1966 upgrade, the N N1I66 diffuser system was also installed. This system was designed and fabricated by General Atomic to their design specifications. specifications. Conversion Conversion to TRlGA TRIGA fuel took place in 1969, and the auxiliaries for pUlsing pulsing operation (transient rod and drive) were designed and built by General Atomic to their operation their specifications. specifications. All building modifications modifications and equipment additions were in conformance conformance with the building codes in existence existence at that time. 3.2 3.2 Meteorological Meteorological Damage There are no design criteria criteria for the protection protection of facility structures structures from meteorological conditions conditions except that all facility structures were constructed constructed to applicable building codes codes in existence existence at the time. The Reactor Laboratory Laboratory has endured approximately approximately 50 years of Wisconsin Wisconsin weather weather with no meteorological meteorological damage. Furthermore, Furthermore, no facility structures structures are assumed assumed to be operable operable in this SAR for the mitigation of any accident accident (see Chapter Chapter 13, 13, Accident Analysis). Analysis) .
- UWNR UWNR Safety Analysis Report Rev. 2 3-1 3-1 2008 Sept. 2008
3.3 3.3 There Water Damage There are no design criteria criteria for the protection protection of facility structures, structures, systems systems and components water damage. The possibility of flooding due to the lake system of Dane County insignificant due to the distance of the Reactor Laboratory components from from County is considered considered Mendota flood plain as Laboratory to the Lake Mendota
- described in Chapter 2, section described section 2.4, Hydrology. Furthermore, Furthermore, no facility structures, systems and components susceptible susceptible to water damage are assumed assumed to be operable in this SAR for the accident (see Chapter mitigation of any accident 13, Accident Chapter 13, Accident Analysis).
3.4 Seismic Damage protection of facility structures, systems and components from There are no design criteria for the protection seismic damage except that all facility structures structures were constructed constructed to applicable applicable building building codes codes in in existence existence at the time. The probability of the reactor probability of a seismic event in the vicinity ofthe reactor site is considered insignificant due to the stable regional considered insignificant regional geology (see Chapter 2, 2, section 2.5, 2.5, Geology, Seismology and Geotechnical Seismology Engineering). Furthermore, Geotechnical Engineering). Furthermore, no facility structures, systems and components, including the reactor pool, susceptible components, damage are assumed to be susceptible to seismic damage be operable in this SAR for the mitigation operable mitigation of any accident (see Chapter 13, 13, Accident Accident Analysis). 3.5 Systems and Components construction of At the time of original construction the Reactor Laboratory, ofthe design bases were not provided by Laboratory, design
- General Electric Electric for facility systems and components. With the upgrade to TRIGA TRIGA and TRIGA-TRIGA-accident analyses, including NUREG/CR FLIP fuel, accident NUREG/CR 2387 and Chapter 13, 13, show that by the of TRIGA fuel, reliance upon other systems, structures design ofTRIGA components are not necessary structures and components necessary Therefore, with the exception of the fuel, no other facility general public. Therefore, to ensure safety of the general component is assumed to be operable in this SAR for the mitigation structure, system or component mitigation of any accident.
Nevertheless, Nevertheless, experience components over many years of experience gained on facility systems and components of operation have shown these systems to be highly reliable. Descriptions Descriptions of system design and operation of these systems operation discussed in the succeeding chapters of this SAR. systems are discussed 3.6 References References 1..
- 1. NUREG/CR2387, NUREG/CR2387, Credible Credible Accident Accident Analyses TRIGA and TRIGA-Fueled Reactors, Analyses for TRIGA Reactors, Hawley and Kathren, Pacific Northwest Laboratory, April 1982 1982 UWNR Safety Analysis Report Rev. 0 UWNR 3-2 3-2 April 2000
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- 44 REACTOR DESCRIPTION REACTOR DESCRIPTION 4.1 Summary Description Summary Description 4.1.1 Introduction Introduction The reactor was constructed and installed by the Atomic Atomic Power Power Equipment Department Department of the General Electric Electric Company. The present modification modification employs a core composed composed of TRIGA-FLIP ofTRIGA-FLIP fuel supplied supplied by the General Atomic Company.
Initial criticality achieved on March 2266 thth 1961. criticality was achieved 1961. The original maximum steady state power level was 10 kW. Power was increased to 250 kW on December 7 th 1964 and again increased to December 7th the present maximum steady state power level of 1,000 kW on November November 1 4 th 1967. Operation 14th Operation with FLIP fuel began in March March 1974 1974 with a mixed core containing 9 FLIP bundles. In January 1978 an additional additional 6 FLIP bundles were added. In August 1979 the conversion to FLIP fuel core was completed. Figure 4-1 is a pictorial cutaway view of the reactor. The reactor reactor is a heterogeneous pool-type, fueled with TRIGA TRIGA or TRIGA-FLIP TRIGA-FLIP fuel which is cooled by natural convection. The fuel is m, currently enriched in Uranium U235 currently all 70% enriched , although 20% enriched fuel can also be used. Light water acts as both coolant and moderator as well as being a biological biological shield. The core is reflected reflected on two sides by graphite and on two sides by water, the water-reflected water-reflected areas being
- utilized as irradiation facility locations. The top and bottom reflector region is partially graphite and partially partially water.
A 7-by-9 grid, surrounded by a core box, positions fuel, reflectors, control elements, and irradiation facilities. Core reactivity is changed changed and controlled controlled by three shim safety blades, a regulating blade, and a transient control rod. All control elements elements move vertically vertically in shrouds positioned positioned in the core box or inside a fixed tube as is the case of the transient suspension transient rod. A suspension frame supports the grid box and control element drive mechanisms. The suspension suspension frame, in turn, tum, is supported by the reactor bridge. Cold, clean core excess excess reactivity in the present operational operational core is about 4.3 % Ak/k. ilklk. Control elements (control blades and the transient rod) provide a shutdown margin of about 4.2 % ilklk Ak/k..
- UWNR Safety Analysis Report Rev. 2 4-1 2008 Sept. 2008
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*
- Figure Figure4-1 4-1 University UniversityofofWisconsin WisconsinNuclear NuclearReactor Reactor(UWNR)
(UWNR) UWNR UWNRSafety SafetyAnalysis Analysis Report ReportRev. Rev. 00 4-2 4-2 April April2000 2000
- 4.1.2 Summary of Reactor Reactor Data Responsible Organization Organization The University Wisconsin University of Wisconsin Location Madison, Wisconsin Wisconsin Purpose Teaching Teaching and Research Research Fuel Type TRIGA Hydride in 4 element clusters
- Number of elements in standard 1000 1000 kW k W core Control Safety elements Three vertical blades blades Regulating-servo element Regulating-servo element One vertical blade
- Transient Transient control One rod Experimental Facilities Experimental Thermal Thermal Column One, 40-inch square square graphite, dO th =
= 2 x 10 nv 8
Beam Beam Ports Four, 6-inch diameter 4ý th == 1 - 33 x 10
101010 nv at shield side of shutter; about 8 x 1011 lOll nv at core end of port Pneumatic tube One, 2-inch (sample size 1-1/4 inch diameter by 5-1/2 inches long), d4 th = 10 12 nv
= 55 x 1012 Thermal neutron fluxes for isotope production production include include the above, plus large irradiation irradiation spaces spaces outside outside the core with thermal thermal neutron fluxes of around 1.3 x 1013 nv.
nv. Reactor Materials Reactor Fuel - moderator material U-Zr H1L6.6
- UWNR Safety Analysis Analysis Report Report Rev. 2 4-3 Sept. 2008 2008
- U235 U 235 enrichment 70%
70% U235
-
235 content/element content/element (average) (average) Burnable poison 11.5wt.% 1'qatural Erbium
.5wt.% Natural Cladding sta inless steel 20 mil stainless Reflector Reflector an I graphite Water and Coolant wat er Light water Control Boral && sstainless tainless steel; borated graphite for jrod transient rod Structural material Aluminur Aluminumn Shield Concrete and water Dimensions Pool Standard 1000 kW Core Grid box 8 x 12 x 227-112 27-1/2 ft. deep 15 x 17 x 15 inches high 9 x 7 array arra y of 3 inch modules modules
- Fuel element element Diameter Diameter Nuclear Length Active Active Length characteristics
--
Nuclear characteristics 1 MW Steady State: Maximum Maximum thermal thermal neutron flux 3.2 xX 10"3 3.2 10 13 nv Maximum Maximum fast neutron flux 3.0 13 nv 1013 3.0 xX 10 UWNR Safety Safety Analysis Analysis Report Report Rev. 0 4-4 4-4 2000 . April 2000
*
- 1000 MW Pulse Maximum thermal neutron neutron flux Maximum fast neutron flux 3.2 Xx 10 16 nv 3.0 x 1016 nv Core Loading (Standard 1000 kkW W core) *
- 11 Operating excess reactivity -4.3% AKeff Reactivity in control blades ~7.1% ~Keff
-7.1% AKeff Average prompt temperature temperature coefficient of reactivity -1.26 0- 4 ~K;oc
-1.26 x 110-4 AK/°C Void coefficient ofof 4
reactivity -.2 x 10-10-4 SK/% AK*% void Prompt neutron lifetime 10.55 second 2.4 x 10-Effective delayed Effective delayed
- neutron fraction 0.007 0.007 4.1.3 Experimental Facilities Experimental Facilities Facilities are provided to permit use of radiations from the reactor in experimental experimental work without endangering personnel. These facilities include three hydraulic irradiation facilities ("whales"),
endangering ("whales"), four beam ports, one thermal column, and a pneumatic pneumatic transfer ("rabbit"). transfer system ("rabbit"). 4.1.3.1 Hydraulic Irradiation Facility (Whale) Hydraulic Irradiation Aluminum pipes of Aluminum 2-7/16" internal diameter extend from approximately of2-7116" approximately 18" 18" below the pool surface to grid box positions on the periphery of the core. These pipes draw sample bottles made polyethylene down and position of polyethylene approximately at the center line of the fuel. Two sample position them approximately sample containers can be loaded in each each tube. The addition addition of a second second sample bottle, however, causes Neutron Fluxes in these positions the natural rotation of the first bottle to stop. Thermal Neutron positions are 13 approximately 10 approximately 1013 nv. other reflector regions surrounding the core. Irradiation conducted in the other Irradiations are also conducted Irradiation irradiations can also be conducted baskets may be inserted in any vacant grid position, and irradiations conducted outside the grid box in specially fabricated enclosures. enclosures .
- UWNR Safety Analysis Report UWNR Report Rev. 2 4-5 2008 Sept. 2008
4.1.3.2 Thermal Column Column
- 4.1.3.2 Thennal thermal column The thennal graphite-filled horizontal column is a graphite-filled penetration through the biological horizontal penetration biological shield provides neutrons in the thennal which provides thermal energy range (about 0.025 eV)
(about 0.025 eV) for irradiation experiments. experiments. The column, which is about 8 8 feet long, is filled with about 66 feet of graphite. A about A experimental air chamber (40" x 40" xX 24") between small experimental between the face of graphite and the of the graphite thermal column door has conduits thennal connections (air, water, electricity) conduits for service connections electricity) to the biological shield compensated ion chambers for the safety and 10gN shield face. The compensated instrumentation channels logN instrumentation channels are are located thermal column. located in the thennal Personnel in the building are protected Personnel protected against gamma radiation from the column column by by a dense dense concrete door which closes the column concrete biological shield. The door moves on tracks set column at the biological set perpendicular to the shield face. concrete floor perpendicular into the concrete 4.1.3.3 Beam Ports Four 6-inch penetrate the shield and provide fluxes of both fast and thermal 6-inch beam ports penetrate thennal neutrons neutrons experimental use. The for experimental The ports are air filled tubes, welded shut at the core core ends and provided with water-tight water-tight covers on the outer ends. The portionsportions of the ports within the pool are made made of portions within the shield are steel. aluminum, while the portions A shutter assembly, A assembly, made of lead encased encased in aluminum, is opened opened for irradiations by by a lifting When closed, the shutter shields against gamma device. When gamma rays from the shut-down core, allowing allowing
- experiments to be loaded and unloaded without excessive experiments excessive radiation radiation exposure to personnel.
personnel. Shielding plugs are installed in the outer end of each port. The plugs, made of dense concrete concrete in in aluminum casings, have spiral conduits for passage of instrument aluminum instrument leads. 4.1.3.4 4.1.3.4 Pneumatic Tube Pneumatic Tube A pneumatic tube system conveys samples from a basement room to an irradiation position A core. The "rabbits" used in the system will convey samples beside the core. samples up to 1-114 1-1/4 inches diameter 5-1/2 inches long. The diameter and 5-112 The system operates closed loop with carbon operates as a closed cover carbon dioxide cover 441 generation of Ar ' activity. gas limiting generation activity. pneumatic tube is restricted reactivity effect from a sample in the pneumatic The reactivity restricted to less than 0.2% 0.2% p.p. Tests run with water samples indicate water and cadmium samples indicate that sample reactivity effects will nonnally reactivity effects normally be less 0.0 1% than 0.01 % p.
- p. Static reactivity measurements will be run for samples of fissionable material reactivity measurements material or particularly strong absorbers particularly absorbers such as some of the rare earths. earths.
4.2 4.2 Reactor Core Reactor Core operated with either The reactor may be operated either standard (20%(20% enriched) enriched) or or FLIP FLIP (70% (70% enriched) enriched) fuel as
- described 4.2. 1. In addition, mixed cores containing described in section 4.2.1. containing fuel of both types may be loaded loaded...
UXVNR UWNR Safety Analysis Report Rev. 2 46Sp.20 4-6 4-6 2008 Sept. 2008
Since funding is not presently available for replacing the FLIP fuel with LEU with burnable presently available
- poison, the core that will usually be operated operated is composed only of FLIP fuel in order to maintain the radiation levels of this fuel above the point at which it is self-protecting. The timing of funding for LEU fuel is unknown at present. If funding for converting standard and FLIP cores before loading a new it may be necessary to revert to cores of mixed standard core in order to assure that less than a formula quantity of HEU (as defined in 10 CFR 73.2) ofHEU of converting the entire core is received new becomes becomes non-self-protecting. replacement is received a few fuel bundles at a non-self-protecting. If funding for LEU replacement further analysis of cores standard/FLIP core may not be required, but further time, reversion to a mixed standard/FLIP of mixtures of FLIP and the new LEU fuel would be required. Such cores are not considered considered in this SAR.
The use of the reactor as a training and research tool requires flexibility of core arrangement. arrangements are subject, however, to the following criteria: These arrangements
- a. A mixed core must contain at least 9 FLIP bundles.
- b. Any FLIP fuel must be located central contiguous region.
located in a central
- c. The core must be a close packed array except for single fuel element (not fuel bundle) positions or grid positions on the core periphery.
periphery.
- d. Calculations indicate that operation will be within safety limits on power power generation per element and fuel temperature.
generation temperature.
- UWNR Safety Analysis Report Rev. 2 4 4-7 2008 Sept. 2008
- 4.2.1 Reactor Fuel The fuel is of the TRIGA TRIGA four-element bundle type developed to provide a simple means of of converting reactors converting reactors using flat-plate fuel to TRIGA reactors. A variant bundle, called a three-element three-element bundle, has only three fuel elements elements installed; the fourth space is used for an aluminum control rod guide tube or an instrumented instrumented fuel element. Figure 4-2 shows a four-element four-element bundle, a three-element three-element bundle containing a control containing control rod guide tube, and a three-element three-element bundle bundle containing an instrumented fuel element element (the conduit for the thermocouple thermocouple leads is shown cut short) short)....
- UWNR Safety Analysis Report Rev. 2 Figure Figure 4-2 4-8 4-2 Fuel Bundles Bundles Sept. 2008
- The four-element four-element bundle (Figure 4-3) consists of bottom adapter, top adapter, and four TRIGA
- elements. The bottom adapter of the bundle fits the existing grid plate as did the original flat-plate fuel elements. The end fittings on individual adapter until a flange on the element of the fuel elements.
elements. A sliding individual TRIGA TRIGA elements are threaded into the bottom element seats firmly against the adapter, providing type support. The top adapter serves both as a handling fitting and as a spacer providing rigid cantilever-spacer for the upper ends sliding fit between this adapter and the fuel element end fittings allows for differential differential expansion expansion of the elements. This top fitting can be removed with remote handling tools to disassemble disassemble the bundles for replacement replacement of individual fuel elements or for shipping spent elements for reprocessing. The individual fuel elements (Figure 4-4) are quite similar to the TRIGA elements used on on reactors using the standard TRIGA grid plates. The differences TRIGA reactor~st~mdard differences are (1) reduction of of diameter from _ inches to maintain the proper metal-to-water metal-to-water ratio in this core; (2) the bottom end fixture is threaded; threaded; (3) flats on the stainless stainless steel bottom end fixture provide wrench surfaces disassembly without stressing the cladding; and (4) the top end fixture is modified surfaces for disassembly to allow the top end fitting to be locked in place. The TRIGA elements elements used at UWNR are of two types, standard and FLIP. Both have outside dimensions, clad thickness, and construction as shown shown in Figure 4-4. The two types differ as shown in the following table: Design Parameter Parameter Standard Fuel FLIP Fuel
- Fuel moderator material U235 23S enrichment enrichment U-Zr H 20%
HI.7 1 .7 U-Zr H1.6 70% 70% H1.6 U235 23S content/element content/element Burnable (average) (average) Burnable Poison
*None
*Natural erbium content Erbium content --- %
1.5 wt % The FLIP fuel was designed to extend the lifetime lifetime of TRIGA fuel, and was used in step-wise step-wise additions of fuel to the University of Wisconsin NuclearNuclear Reactor. The reactor is currently operating with a core consisting entirely entirely of FLIP fuel. Fuel bundles contain only one type of fuel, and the top adapters for FLIP fuel bundles are marked (by notches notches machined into the top of the top adapter) to facilitate identification identification during underwater underwater fuel handling.
- UWNR UWNR Safety Analysis Report Rev. 0 4-9 April 2000
*
*
- Figure 4-3 Four-element Four-element Bundle Assembly Assembly UWNR Safety Analysis Report Rev. 0 4-10 4-10 April 2000
- Figure Figure 4-4 Fuel Element Construction Element Construction UWNR Safety Analysis Report Rev. 0 4-11 4-11 April 2000
Figure 4-5 shows an instrumented instrumented element. This element is fitted with three thennocouples. thermocouples. At
- least one such element (inserted (inserted into the vacant position of a three-element three-element bundle) is included in every every core. The sensing sensing tips in the thermocouples thennocouples are located at the vertical centerline centerline of the fuel section and one inch above and below the centerline. thermocouple leads pass through a centerline. The thennocouple seal in a stainless steel tube which provides provides a water-tight water-tight conduit carrying carrying the lead-out wires wires above the surface of the pool water.
- Figure Figure 4-5 Instrumented UWNR Safety Analysis Report Instrumented Fuel Element Report Rev. 0 4-12 4-12 April 2000
- 4.2.2 Control Elements
- Both blade and rod shaped control elements are used.
4.2.2.1 Control Blade Shrouds and Guide Tubes Each blade type control element, both 27.4"- safety and regulating, is guided throughout shown in Figure its travel by a shroud as shown 4-6. The shroud consists of two thin separated aluminum plates 38 inches high, separated 1/8-inch aluminum spacers to provide a 1I8-inch by aluminum water annulus. The shrouds can be removed, if necessary, by use of a grapple grapple
'1--
hook. Small flow holes at the bottom of of I I I I II minimize the effect of viscous the shroud minimize II II I damping on the scram time. damping Ii I II II II I ~I I II I II II I II I I I 01 I II I II
- control elements are guided by Rod shaped control I I II
-
co I I II a guide tube as shown in one of the three- -Y) (Y)
- -~-~-
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I II I element bundles of Figure 4-2. Holes element Holes I II Ii I II II I drilled in the sides of the guide tube allow allow 1/ IIII II I I I II for water displacement when the control I: Ii II 0 rod is fired out during pulsing operation or II , U I Ii II E-dropped in response to a scram condition. II II II 0 I I' II II I"I
.----- - ---.,->t---- ;;.==:i"--
Figure 4-6 Control Blade Blade Shroud
- UWNR Safety Analysis Report Rev. 2 UWNR 4-13 4-13 2008 Sept. 2008
1 4.2.2.2 Reactor Figure Safety Blades Reactor control for startup and shutdown is accomplished Figure 4-7, with a total shutdown worth between 6.9 and 11 is boral sheet (boron carbide carbide and aluminum accomplished by three blade-type aluminum sandwiched safety blade is 40.5 inches long. When a blade safety sandwiched between blade-type control elements, 11 per cent AD. KCff. between aluminum Keff* The poison section aluminum side plates). Each blade is full in the bottom of the blade overlaps section Each overlaps the
- bottom of the active fuel by 1.5 inches.
YP"
/* BORAL SHEET 1/8" 1/8" CLADDING - - - + - -
ALUMINUM CLADDING" TOP OF ACTIVE FUEL ELEIvJENTS -----+---- - - - - - - - - e ,REACTOR T'C'TTOi; OF CORE
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Figure 4-7 Shim/Safety Shim/Safety Blade 1 4,2.2.3 4.2.2.3 , Regulating Blade The regulating blade, Figure 4-8 (shown (shown upside upside down for ease in reading the dimensions), dimensions), is a stainless-steel stainless-steel sheet with curls on the vertical edges, about 11 inches wide and 40 inches long, supported and guidedguided in the same manner manner as the safety blades. It is used to compensatecompensate for small changes of reactivity during normal reactor operation operation and may be actuated actuated by a servo-control servo-control channel. UWNR Safety Analysis Report Rev. 2 4-14 4-14 Sept. 2008 2008
*
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- Figure Figure 4-8 4-8 Regulating Regulating Blade 4.2.2.4 Transient Control Rod Transient The transient transient control rod is boron carbide carbide or borated graphite contained contained in a 1.25 inch diameter stainless 1.25 stainless steel or aluminum aluminum tube (Figure (Figure 4-9).
4-9). The approximately 15 poison section is approximately 15 inches inches long. This rod is guided laterally laterally by the aluminum guide tube in a special three-element by three-element fuel bundle. A hold-~ A hold-down tube extends extends from this guide guide tube to the top of the reactor reactor structure structure and holds the three-element three-element bundlebundle in place during during transient transient rod movement. ~ Figure 4-9 Figure 4-9 Transient Control Rod
- UWNR UWNR Safety Analysis Analysis Report Report Rev. 2 4-15 Sept. 2008 2008
- 4.2.3 Moderator and Reflector Neutron Moderator Reflector Pool water serves as moderator moderator for the core and as reflector above, below, and on those sides of of the core not provided with graphite reflectors reflectors or special reflector elements designed to conditioncondition the quality of a beam beam being extracted extracted from the core. (Individual (Individual fuel elements contain contain an internal 3.5 inch long graphite end reflector above and below the fueled portion, so the core top and bottom are actually reflected by a mixture of graphite and water.)
The reflectors are standard standard General Electric Electric Company reflectors furnished at initial startup of UWNR. The ofUWNR. SIDE IrV:li
--.~-
nominal 3-inch square reflector reflector elements are made of AGOT grade graphite clad with aluminum GRAPHITE aluminum (Figure 4- - 10). Reflector element lifting handles 10). Reflector are diagonal diagonal to facilitate identification identification when viewing the core and storage storage racks. 1/16" ALuMINUM CLADDING Special Special reflectors reflectors are the same size as the graphite reflectors; but may consist of solid aluminum, hollow aluminum, or combinations combinations of graphite sections with the center center portion portion replaced replaced with TOP VIEW
- -VIEW
-
*
[§]~ solid aluminum, voids, or gamma absorbers gamma absorbers such as lead or bismuth. IT Such special reflectors are used for irradiation irradiation facilities or to adjust the mix k of thermal, epithermal, epithermal, and fast neutrons transmitted transmitted to experimental experimental I--- 3" NON 3" NOM --I facilities. Figure Reflector Element Figure 4-10 Graphite Reflector Element UWNR Safety Analysis Report Rev. 0 4-16 4-16 April 2000
- 0
- 4.2.4 Neutron Startup Source The neutron source is a _ radium-beryllium irradiated to give an output greater than radium-beryllium source irradiated 10' neutrons/second. It is encapsulated in a 0.515 inch diameter 10 7 diameter by 3.10 inch long stainless steel welded cylindrical cylindrical capsule, capsule, which in turn tum contains two 1.25 inch long welded stainless stainless steel capsules.
The source source fits into a source holder (Figure 4-11). The source source holder, in turn, tum, fits into an irradiation irradiation basket (Figure 4-12) occupying one grid module adjacent to the active core. The source is usually left in for full power operation, and will, with the normal operating cycle, maintain its output of about 10' 107 neutrons/second. If the source is not left in during full power operation, neutron emission rate will decrease over over a long time period to approximately 10 1066 neutrons/second. neutrons/second.
- UWNR Safety Analysis Report Rev. 2 4-17 4-17 2008 Sept. 2008
c UWNR Safety Analysis Report Rev. 2
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- UWNR UWNR Safety Analysis Report Rev. 2 4-19 4-19 Sept. 2008
- 4.2.5 Core Support Structure The core is suspended from an all-aluminum all-aluminum frame, Figure 4-13, which extends from the grid box to a height of about one foot above the pool surface. One of the hollow comer comer posts of the suspension frame serves as a guide for the gamma chamber used in pulsing operation. The other other three comer comer posts may also be used to position detectors in positions above the core.
The reactor bridge (mounted (mounted over the pool) supports the core suspension suspension frame. The all-steel, all-steel, prefabricated prefabricated bridge was bolted together in the field and aligned with shims. A locating locating plate, made ofO.5-inch of 0.5-inch steel, spans the upper end of the suspension suspension frame. It is bolted to the bridge and aligns the four control blade drive mechanisms and the transient transient rod drive with the core. The five mechanisms mechanisms work through individual clearance clearance holes, each mechanism being secured secured to the locating plate. The plate and mechanisms mechanisms are not removable as a unit to prevent accidental accidental withdrawal of the control elements. The fission counter drive is mounted on a portion portion of the hand railing support structure. Four 4-inch square 6061 aluminum suspension tubes (0.25 inch wall thickness) extend from the bridge bridge to the grid box, and support the grid box by bolted connections. The aluminum grid box, Figure Figure 4-14, encloses encloses and supports the 6-inch thick cast aluminum grid plate which is machined to locate and support the control element element shrouds and bottom end fittings of fuel bundles, reflectors reflectors and in-core experimental experimental facilities such as hydraulic irradiation positions and , irradiation baskets. Figure 4-15 shows the grid position designation designation system, location ofof
- experimental facilities and radiation detectors relative to the grid box, and letter and number codes used in later descriptions to identify fuel and reflector reactivity worths.
An aluminum coolant headerheader (not shown in the figures) mates with the bottom of the grid box and forms a transition to the coolant piping originally originally provided provided (but not used) for future use with a forced convection convection cooling system. An opening in the side of the header, 24 inches wide by 12 12 inches high, allows cooling water flow for natural convection. convection. A diffuser pump and jet above the core deflects the cooling water streaming streaming from the core to reduce N 16 I6 activity above the core. reduce activity UWNR Safety Safety Analysis Analysis Report Rev. 2 4-20 4-20 Sept. 2008
*
- Figure Figure 4-13 4-13 Core Core Suspension Suspension
- UWNR UWNR Safety SafetyAnalysis Analysis Report ReportRev.
Rev. 22 4-21 4-21 Sept. Sept. 2008 2008
*
- Figure 4-14 Grid Box and Grid Plate UWNR UWNR Safety Safety Analysis Analysis Report Report Rev. 2 4-22 4-22 2008 Sept. 2008
*
- Figure 4-15 Grid Arrangement Arrangement With Fuel and Reflector See Table 4-1 and 4-2 for an explanation Reflector Codes explanation of coding in the figure above.
- UWNR Safety Analysis Report Rev. 2 4-23 Sept. 2008 2008
4.3 4.3 Reactor Pool Reactor The aluminum-lined deep. The reinforced the pool Pool concrete pool, aluminum-lined concrete reinforced concrete Figure 4-16, pool, Figure 4-16, is 88 feet wide, pooi walls are found in the next section). The pool wide, 12 feet long, and 27 1/2 feet 27 112 biological shield (further details on concrete pool walls also serve as the biological pool is penetrated by experimental penetrated by experimental ports. on
- Piping systems connecting connecting with the pool are discussed discussed in detail in Chapter 5. 5. In summary, all all26 piping connections connections are built to preclude accidental26 preclude accidental pool water by loss of pooi by failure of components components located outside outside of the pool. When When possible, possible, pipes enter and leave the pool above the water surface surface cooling system, diffuser system, and (primary cooling and hydraulic irradiation tube water system) and are-hydraulic irradiation are equipped equipped with passive siphon with passive siphon breakers breakers that that prevent loss of more than a few inchesinches of water even in the event of a pipe break or system system mis-operation.
Two 88 inch aluminum aluminum pipes intended intended for use in a forced-convection forced-convection cooling system were imbedd were imbedded in the concrete at initial construction. (~ee Figure construction. (See 5-2,
- especially the note concerning 5-2, especially concerning anti-siphon anti-siphon loop). One of these pipes penetrates penetrates the pool pooi wall about 14 feet below the pool curb, but is closed about closed-with flanges on both the inside and outside outside ends, preventing preventing loss of pool water unless unless both flanges fail. The other 8 inch pipe was looped inside the fail. other 8 inch pipe The and concrete wassiphon inside -the loopedbreaker concrete and equipped equipped with with aa siphon breaker that that extends from the top of the looploop to above the pool curb. One end of the loop is flanged closed outside outside the shield, while the other end vertically penetrates the bottom of the pool within the penetrates coolant coolant header header (below the grid box). If If the outer outer flange of this pipe were to fail, this pipe could could___
drain the pool to 14 feet below the pool curb. At Fiue41RacoPolndSed Safety Analysis UWNR Safety Report Rev. 2 Analysis Report 424Sp.08 4-24 4-24 2008 Sept. 2008
*
- purification system supply pipe penetrates the pool 38 inches below the The pool makeup and purification pool surface, thus limiting water discharge pipe water loss to that level reduction upon failure. The discharge from this system is discussed discussed in the paragraph immediately immediately above.
mid-core level. The beam ports are operated with a penetrate the pool wall at mid-core The 4 beam ports penetrate watertight flange on the outside end which will prevent watertight in-pool portion of the prevent leakage should the in-p'ool beam port fail. The in-pool portion of the tubes are aluminum pipes with welded end closure and bolted flanged connection to the beambeam port shutter assembly. The four beam ports have a common drain system, but the discharge valve for the drain system is maintained maintained closed during connections which connect to the Beam Port and operation. The beam ports also have vent connections Thermal Column Ventilation system (see Chapter 9, Section 9.1). The vent connections are Column Ventilation equipped with valves (normally kept open for ventilation flow), but which may maybe be closed if a beam port leak occurs. Finally, each beam port vent has a check valve which which allows air flow only out of the vent, thus preventing pressure differences differences between between the beam ports from causing circulation circulation between between beam ports. ports .
- penetrates the pool wall. It is of welded aluminum The thermal column case also penetrates valves or flanges which could be opened to drain the pool.
and has no valves aluminum construction construction The pool water is kept within the following limits: 0F cooling water inlet) Temperature (at Core cooling Temperature <130
<130°F Resistivity 10'5 Ohm -- cm
>2 x 10 cm Radioactivity Radioactivity <10 CFR Part 20 Appendix B Table 33 values radioisotopes with >24 hour half-life for radioisotopes The reactor core is cooled by natural convection of pool water through the core. The 130°F 130'F temperature limit is imposed by demineralizer temperature demineralizer resin tolerance and by humidity control
- considerations.
considerations . UWNR Safety Analysis Report Rev. 2 UWNR 4-25 2008 Sept. 2008
The resistivity limit is set to reduce corrosion effects, extending extending the expected expected lifetime of the fuel
- elements and controlling controlling water radioactivity. Routine checks of resistivity are made to determine the necessity of regenerating the demineralizer.
demineralizer. The radioactivity radioactivity of the pool water is continuously monitored by an area monitor station located near the demineralizer. Should the pool water reach the activity limit above, the reading on this area monitor will increase during periods periods when the reactor reactor is not operating. In addition, water water samples are routinely analyzed for activity by other methods which give a more exact identification of quantity and type of activity present. No problem has beenbeen experienced experienced in maintaining maintaining pool water radioactivity radioactivity below the indicated I* limits in nearly nearly 50 years of operation. 4.4 Biological Shield Biological Shield The reactor is shielded by concrete and water (See Figure 4-16). At normal pool level the core is covered by 20 feet of water. The shield at core level consists of about 3 feet of water plus 8 feet (9 feet on thermal column side) of ordinary ordinary concrete. Denser concrete is used in the thermal column door and beam port port plugs. Calculations and measurements Calculations measurements of radiation radiation levels for 1000 1000 16 l6 kW operation are (excepting N activity) discussed below: Surface of shield, excepting Pool surface (leakage and thermal excepting beam port and thermal column openings 16 16 (leakage radiation) (No N )) less than 15 mremlhr.
"Hot spots" - measurements have shown thermal column. Measurements mrem/hr.
mrem/hr. openings - less than 1.5 mremlhr. shown that higher radiation levels exist around the beam ports Measurements of the maximum radiation levels at these "hot spots" at 1000 1000
- kW k Wareare about 10 mrem/hr mremlhr around the beam ports and 40 mrem/hr mremlhr at the hottest spot around the thermal column door. The dose one foot away from the hot spots is about 5 mrem/hr. mremlhr.
Since the third and fourth-floor classrooms and offices, and fifth-floor mechanical mechanical room, are above the level of the pool curb, an analysis ofthe of the effect of complete water loss on persons in these areas was performed performed and results included included in the updated Emergency Emergency Plan. The computer computer code MCNP5 was used to model the dose rate, as described in section 13.1.3.2. 13.1.3.2. Since the biological shield does not offer any shielding to the central wing third floor classrooms, the dose in these classrooms would be greater than any location location on the fourth or fifth floors. The building building evacuation alarm would evacuate evacuation evacuate people from these areas before the pool was completely completely drained, such that the total integrated integrated dose received by evacuating evacuating members members of the public would be about 13 mrem. UWNR UWNR Safety Safety Analysis Analysis Report Rev. 2 4-26 2008 Sept. 2008
- Pool Surface Radiation Levels - N1 N I66 Activity
- 4.4.1 The radiation level due to N N"I66 activity at the pool surface directly above the core when operatingoperating at 1000 kW would range from 80 to 220 mrem/hr mremlhr ifno if no N-16 control system were in operation operation (variability is due to changes (variability changes in surface flow patterns). The N-16 N -16 diffuser system is normally in operation, however, reducing the dose rates at the pool surface to 2-4 mremlhour at the pool surface. These radiation radiation levels are low enough that no hazard will exist exist to personnel personnel outside the Reactor Laboratory Reactor Laboratory or in normally occupied levels within the Reactor Reactor Laboratory. Radiation Radiation levels on the walkway surrounding surrounding the pool are around mrem/hr while the reactor is operating around 20 mremlhr at 1000 kW without the diffuser operating operating and <0.5 mrem/hour mremlhour with the diffuser operating.
All of the Reactor Laboratory outside of the console area is posted as a radiation area and a radioactive materials area. A cable and switch arrangementarrangement is positioned on the north stairway stairway to the pool surface so that an alarm will be sounded should entry to that area be made while the reactor is operating, thus assuring that personnel personnel will not enter the area without knowledge of the reactor operator. The south stairway, leading leading from the console console area to the pool surfacesurface does not have a cable and switch arrangement arrangement as does the north stairway. Access stairway. Access to these stairs is gained only through through the console console area and is well monitored. No difficulty has been experienced experienced in maintaining radiation radiation doses to individuals individuals well below below those doses permitted permitted in 10 CFR 20.
- 4.4.2 Heating Heating Effects in Shield and Thermal Column Column Heating effects caused by absorption of gamma gamma radiation radiation and fast neutrons neutrons are within allowable allowable limits. For all calculations, calculations, it was assumed that the pool water was at the 130° 130' temperature limit, and the reactor was operated continuously continuously at 1.5 MW.
The heating in the concrete shield shield is approximately approximately 20% of the maximum maximum suggested suggested by by Rockwell'. Rockwell I . Analysis of the heating rate in the lead shield for the thermal column indicates that the maximum maximum temperature temperature of the lead will be less than 217°F. Calculation of the graphite temperature in the thermal column indicates temperature indicates a maximum of 244°F. 4.5 Nuclear Design 4.5.1 Normal Operating Operating Conditions Conditions and Reactor Core Physics Paremeters Note: NUREG-1537 NUREG-1537 specifies separate sections for "Normal Operating Conditions" Conditions" and "Reactor "Reactor Core Physics Parameters." Parameters." These two sectionssections are combined to enable concise inclusion of the measured measured core parameters parameters of the several several cores which have been operated under the license.
- UWNR Safety Analysis Report Rev. 2 UWNR 4-27 Sept. 2008
1 4.5.1.1 Core Arrangements
- Arrangements The use of the reactor as a training and research research tool requires flexibility of core arrangement.
arrangement. Permitted arrangements Permitted arrangements are subject, however, to the following criteria:
- a. A mixed core must contain at least 9 FLIP fuel bundles (clusters)
- b. FLIP fuel must be located located in a central contiguous region region
- c. The core must be a close packed packed array except for single fuel element positions or or fuel bundle positions on the core periphery
- d. Calculations indicate that operation of a specific core will be within technical Calculations specification limits on power generation specification generation per element and fuel temperature.
When the Safety Analysis Report When Report for converting converting UWNR from flat-plate to TRIGA fuel was written, expected performance performance was based based on computations and on the behavior of a "prototype""prototype" TRIGA Mark III III reactor, the Torrey Pines Reactor Reactor at General Atomics. The prototype reactor used individual TRIGA fuel elements in a right circular cylindricalcylindrical array typical of TRIGA TRIGA reactors; UWNR uses four-element bundles in a rectangular rectangular arrangement arrangement in the grid box provided for the original flat-plate fuel. The uranium loading in the prototype prototype was 8 wt% uranium, while UWNR has a uranium loading of 8.5 wt%. Both the prototype prototype and initial UWNR UWNR TRIGA cores TRIGA cores had stainless steel clad, and both used 20% enriched 20% enriched uranium. The heat transfer characteristics characteristics were quite similar, although although the diameter of the clad for UWNR was slightly smaller to fit to the grid box array spacing. UWNR smaller UWNR also differed by having shrouds dividing the core box into three regions. These shrouds guide control blades, blades, but also introduce introduce water water gaps within the core lattice. The prototype was operated operated for many years at steady-state steady-state power levels up to 1500 kW and thousands of pulses up to 6000 MW. In this report, although the prototype performance characteristics characteristics are indicated performance compared to UWNR, most of the information is indicated and sometimes compared
- based on the measured performance performance of the cores which are currently operable under the present license license and technical specifications.
The current core is an all-FLIP (stainless steel clad) configuration consisting of23 of 23 FLIP fuel bundles. Cores with 9 and 15 FLIP fuel bundles also have been operated for significant significant times, and have been thoroughly tested for conformance technical specifications conformance to technical specifications and the predictions and descriptions in this and the previous SafetySafety Analysis Report. It is planned planned that the all-FLIP core will continue to be the operating core until funding (and analysis) for refueling with LEU is is completed, at which time this SAR will be amended. It may become become necessary necessary to revert to the 15- or 9-bundle FLIP cores before refueling, however, in order to maintain less than a formula 15-quantity of HEU fuel at non-self protecting protecting levels during the return of the HEU to DOE. UWNR Safety Analysis Report Rev. 2 4-28 2008 Sept. 2008
- Standard TRIG 4.5.1.2 Standard TRIGA A fuel cores2
- This core, shown This fuel bumup fuel shown in Figure Figure 4-17, and aa succeeding core enlarged to 30 fuel bundles because of burnup was operated from November 1967 until March 1974 when it was shut down.
Measured core parameters parameters for the initial 25 bundle version of the core are presented below. below. of variations in the peripheral reflector configuration were included in this series of cores, Several variations including up to 18 reflector elements around the periphery, and some cores with special voided-center or Bismuth-center reflectors. Core Designation A25-RIO A25-RI0 Excess reactivity 4.82 % Pp Shutdown Margin % P 5.17 % p Transient Rod worth 2.10 % % Pp reactivity defect 2.95 % P FP reactivity p Peak pulse power 1930 MW Prompt neutron lifetime 42E-6 sec
- Fig Figure 4-17 Standard TRIGA Fuel Core
- UWNR UWNR Safety Analysis ReportReport Rev. 2 4-29 4-29 Sept. 2008 2008
Themeasured The measured fuel fuel temperatures temperatures (Figure (Figure 4-18) 4-18) inin the the UWNR UWNR standard standard TRIGA TRlGA core core almost almost
- matched those in the prototype, although this could have differed significantly, matched those in the prototype, although this could have differed significantly, depending upon depending upon instrumented instrumented element element placement placement inin the the two two cores.
cores. The The UWNR UWNR instrumented instrumented element element was was located located near the core center in grid position near the core center in grid position D4NW D4NW as as indicated indicated in in Figure Figure 4-17. 4-17. I S1.Ii' :::.:I.,>~.. 1. I
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- REACTO~ POWER, REACTOR POWER, KILOWATTS KILOWATTS Figure Figure 4-18 4-18 Fuel Fuel Temperature-Temperature- Standard Standard FuelFuel ,
UWNR UWNR Safety Safety Analysis Analysis Report Report Rev. Rev. 22 4-30 4-30 Sept. 2008 2008
*
- When the reactivity loss from power operation for UWNR with the original TRIGA compared prototype (Figure 4-19) it is apparent that the power compared to the prototype significantly larger than for the prototype. In both cases, the core had been pulsed a significant number of times before the temperature clad stretching in pulsing. The large temperature measurements were made, so the difference is not from large loss was considered power defect TRIGA core is defect for UWNR is considered to be a function of the different core significant is from geometry and reflection making the leakage change more drastically with temperature in UWNR.
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~+~+LU 0oo 200 4oo 600 8oo 0ooo 1200 1400 160c Reactor Power, Power, Kilowatts Kilowatts Figure 4-19 Power Defect vs. Power- Standard TRIGA TRIGA Core When pulsing behaviorbehavior of the two cores was compared, compared, other differences expected and differences were expected
.found. The pulsing behavior differed from the typical TRIGA core primarily due to the water control blade shrouds and the graphite reflector, both of which increased neutron gaps in the control lifetime, resulting in longer longer periods for the same pulsed reactivity insertion, and thus broader broader pulses. Later graphs, Figure Figure 4-35 through Figure 4-37, compare compare the pulsing behaviorbehavior of this and all of the other pulsingpUlsing cores with that of the prototype TRIGA TRIGA reactor. Note that the pulsed reactivity addition addition limit for the standard TRIG TRIGA A fuel was 2.1 % p, p , instead ofthe of the 1.4 % P p for cores cores containing FLIP fuel. Pulsing behavior differences differences will be discussed later in this report.
report .
- UWNR Safety Analysis Report Rev. 2 UWNR 4-31 Sept. 2008 2008
- 1 4.5.1.2.1 Reactivity Reactivity Effects In Standard Standard Fuel Cores The reactivity effect of fuel bundles in Table 4-1 Fuel Bundle Worths in UWNR Cores-different lattice lattice positions have been been all in % pp estimated estimated from measured and calculated calculated Position Water Reflected Reflected Graphite Graphite Reflected Reflected values. The position position codes in Table Table 4-1 shown in Figure 4-15. The are those shown A 2.60 4.0 reactivity value given is the worth of of adding or removing a fuel bundle while the B 1.95 3.46 3.46 remainder remainder of aa.-bundle
-bundle core is already 1.22 2.18 C 2.18 present. The worth of the bundles bundles when added in an approach approach to critical would be C' 1.16 1.15 1.15 radically different, different, since cores cores are loaded loaded as compact cores cores during the loading D 0.77 -
sequence; that is, the fuel loading plan is sequence; E 1.76 2.76 planned planned to assure that the next fuel bundle bundle loaded will have a smaller reactivity effect F 1.85 1.55 1.55 than the bundles previously loaded. Table 4-2 Reflector Element Reactivity Worths
- The reactivity effects of reflector variations Position Replace Replace water Replace water water also have been measured and/or estimated estimated with graphite with air % p and are indicated in Tables Tables 4-2 and 4-3.
4-3. %
%pP First, the worth of both a graphite reflector reflector element and a voided reflector element are 1 1.228 1.228 -0.239
-0.239 indicated indicated relative to a water reflector, reflector, with 2 0.180 -0.198 the position codes being those shown in Figure 4-15 4-15. . 33 0.058 -0.064 4 0.076 -0.083 55 0.115 -0.126 6 0.157 -0.172 7 0.029 -------
UWNR Safety Safety Analysis Report Rev. 2 4-32 2008 Sept. 2008
- changes that affect reflection Next, the effect of other changes of these were reflection are indicated. Most ofthese
- measured in a core of standard TRIGA fuel.
Table 4-3 Condition Reactivity Effect of Reflector Region Changes Result-%p Result-%p Changes Flooding all 4 beam ports Flooding +0.0005 pneumatic tube Flooding pneumatic +0.002
+0.002 Pneumatic tube samples water filled -0.0003 Cadmium filled Dropping fuel bundle on top of +0.5
+0.5 core Adding fuel bundle on side of core +0.77
+0.77
- UWNR Safety Analysis Report Rev. 2 UWNR 4-33 2008 Sept. 2008
4.5.1.3 Cores containing FLIP fuel The longer operating lifetime for FLIP fuel was the major reason for selecting refueling the University of Wisconsin Nuclear Reactor. The higher enrichment coupled with erbium poisoning poisoning provides the longer operating selecting this fuel type for enrichment of FLIP fuel operating lifetime, but it also causes changes
- in operating operating characteristics characteristics relative to standard fuel. The prototype FLIP core was also the Torrey Pines TRIGA TRIGA Mark III fueled with FLIP fuel. The most marked changes from use of FLIP fuel are a reduction of prompt neutron cycle time to about 10E-6 1OE-6 seconds at beginning of core life (20E-6 at end of core life) and a temperature coefficient that is strongly temperature temperature coefficient dependent. (Figure 3-16, page 33 of reference)3 . These data are for the prototype ofreference)3. prototype reactor; values in UWNR were expected expected to and do differ because of the water gap in the control blade shrouds and the graphite reflector, making the neutron neutron lifetime considerably considerably longer in UWNR FLIP cores than it was in the standard TRIGA core.
In addition, addition, the harder harder spectrum spectrum in a FLIP core leads to power peaking in regions near water gaps. This leads, in a compact compact core, to a peaking factor within a FLIP element of 1.43. 1.43. If a large water-filled flux trap is located adjacent adjacent to an element, the peaking factor in the element can increase to 2.65 peak/average peak/average within the cell. Thermal and hydraulic parameters of FLIP fuel remain the same as standard standard fuel. FLIP fuel elements are not mixed with standard elements elements in the same fuel bundle at Wisconsin. Thus, the smallest increment increment of FLIP fuel addition possible will be three FLIP elements elements (in a
- bundle containing the transient rod guide tube). Placing Placing such a bundle in the center of a 5 x 5 array of standard standard TRIGA fuel leads to the highest value of power peaking peaking possible, with resultant resultant, power generation of 31.2 kW in each element. Although no operation operation with this core is is anticipated or desired, other TRlGA anticipated TRIGA reactors reactors have operated with power generation rates at least as high as 32kW 32kW per element.
of less than five FLIP fuel bundles (24 FLIP elements) Addition ofless elements) was not considered considered useful for a full power operating core, since it would not provide sufficient additional additional reactivity to compensate compensate for burnup in the standard elements. Calculations Calculations were performed performed for cores with 1,2,5,9, 1, 2, 5, 9, 15 and 25 FLIP -bundles in central contiguous contiguous regions of the core. All calculations calculations were for a 5 x 5 array array of fuel bundles with the transient rod guide tube in the fuel bundle bundle at grid-position D5. Calculations performed with Calculations were performed a two-dimensional two-dimensional diffusion theory code (Exterminator (Exterminator 2). Standard Standard seven seven group cross sections sections obtained from Gulf-General Gulf-General Atomic were used in the calculations. The accuracy of the calculations calculations was checked by analysis analysis of cores cores with known values ofKeff of Keff and power density. Results calculation of mixed cores and FLIP cores Results on calculation cores were found to be consistent with similar calculations calculations performed elsewhere44 *. Subsequent performed elsewhere Subsequent computations computations using a 3-dimensional 3-dimensional diffusion theory code (DIF-3D) in support of use of LEU fuel agreed well with the results for Exterminator-2 Exterminator-2 for the all-FLIP core. UWNR UWNR Safety Analysis Report Rev. 2 4-34 4-34 Sept. 2008 2008
*
- FLIP fueled cores can experience experience significant significant power peaking which must be considered in considered in permissible fuel arrangements arrangements and setting of limiting safety system settings. The power power produced in individual fuel elements was predicted predicted from the computations computations done for safety analyses. The following table shows both the power density in individual fuel elements elements and the worth of a fuel bundle loaded loaded in a particular particular location (if (if applicable applicable to the condition) condition) for several different core core arrangements arrangements analyzed. See Figure 4-15 for position position descriptions.
Table Table 4-4 Maximum Power Density and Reactivity Worth of Fuel Bundles- Cores Containing FLIP FUEL arrangement (keyed to fuel Core arrangement Power in Maximum Maximum Reactivity Reactivity Effect of Effect of bundle locations locations in Figure Figure 4-15) Element- kW Removing Removing Fuel Bundle in in NOTE: D5 D5 is a 3-element bundle Indicated Position-%p Indicated Position-%p with transient transient rod guide tube 55 FLIP +20 Standard Standard fuel bundles bundles 21.4 21.4 ---- (FLIP in Positions A & & B) Replace Replace FLIP in E5 with H 22O 0 28.3 2.83 9 FLIP +16
+ 16 Standard fuel bundles (FLIP in Positions A, B, and E)
(FLIP 18.1 ----
----
Replace FLIP in D5D5 with H H20 2O 20.0 0.93 Replace FLIP in C5C5 or E5 E5 with H H 220O 25.9 1.69 1.69 Replace FLIP in D4 or D6 with H H20 2O 23.2 0.98
- Replace FLIP in E4, C4, C6, or E6 with 22.3 1.49 1.49 0
H22O
+ 10 Standard 15 FLIP +10 Standard fuel bundles bundles (FLIP in Positions A, B, C, E, &F)
&F) 17.2 17.2 ----
----
Replace FLIP in D505 with H22O0 19.0 19.0 0.87 Replace FLIP in C5C5 or E5 E5 with H HP 20 24.6 24.6 1.65 Full FLIP- 25 FLIP fuel bundles bundles (FLIP in all positions except D)
- 0) 15.5 15.5 ----
Replace FLIP in D5D5 with H H20O 17.2 0.79
. Replace FLIP in C6 or D6 with HP H20 20.0 0.51 Replace FLIP in C5C5 or E5 E5 with H22O 0 22.1 1.42 1.42 Reference to the table above (Table 4-4) shows that power Reference power generation in any individual individual element element is well below 23 KW in all compact compact FLIP fuel arrangements.
arrangements. Further, the presence of a 3-inch square square water gap in the FLIP fuel region will result in power generation generation rates below 23 KW/element KW/element in most of the cores. The initial operational operational mixed core contained nine FLIP fuel bundles (35 elements), and the calculations calculations indicate that flux traps could not be permitted for full power operation operation in this C5, E5, E5, D4, and D6 if the maximum
- arrangement arrangement for locations locations C5, kW/element is to be kept below maximum kW/element below UWNR UWNR Safety Analysis Report Rev. 2 4-35 2008 Sept. 2008
23 kW. 23 kW The combination combination of fuel bundle proximity proximity to control control blade shrouds shrouds and the transient rod
- guide tube causes causes the greatest greatest power power peaking peaking in any of these these cores.
It is also apparent apparent from comparing comparing Tables 4-1 and 4-4 that the reactivity reactivity worth of an individual individual FLIP bundle is lower lower than that of a standard fuel bundle, bundle, even in mixed cores. The Technical Technical Specifications Specifications under under which which the facility has operated since conversion operated since conversion to FLIP fuel required at least nine (9)(9) FLIP fuel bundles bundles (35(35 elements) in a central contiguous contiguous location with no water gaps larger than a single elementelement except except on the core periphery. periphery. As a resul t, the maximum result, power power density in any fuel element 1,000 kW was limited to 18.1 element at 1,000 18.1 kW for any of the cores cores considered. This is approximately approximately 11 11%% higher than the maximum in an all standard fuel core. FLIP-containing cores have been operated Three different FLIP-containing operated at UWVNR. UWNR. Characteristics Characteristics of of each each core, measured measured during the startup startup and acceptance acceptance testing of each each core, are shown in the following sections. During During core test programs, quadrant of each core containing programs, one quadrant containing FLIP fuel was mapped for temperature temperature by by moving an instrumented fuel element into into each unique core position. Interpretation of the fuel measurements Interpretation complicated by measurements was complicated by instrumented instrumented element failures so that measurements were made with different measurements different instrumented elements in different cores as explained instrumented elements explained below. The standard standard fuel temperature temperature measurements measurements were made using two different different instrumented elements, since the original instrumented elements, original standard standard instrumented instrumented element element failed before the 15- 15-bundle FLIP core was tested. There was, however, however, only one standard standard instrumented element in instrumented element the core at a time, so no comparisons comparisons between between the indication of the elements elements in the same same core position are available. least one common common position in the same core position. fail, three different to instrumented FLIP available. Two instrumented the temperature mapping. The individual enable position. However, because different instrumented FLIP elements individual FLIP comparison between because one instrumented instrumented FLIP elements the comparison between the indication available, and both were used in elements were available, FLIP instrumented elements were instrumented elements indication were both of both placed in at the different different instrumented element had all thermocouples elements have been used in the tested cores, thermocouples cores, and at elements widely and widely
- temperatures (as much as 130° varying temperatures 1300 C)C) were were measured when these different different elements elements were placed in the same core position.
position. ThisThis makes interpretation interpretation of the predicted predicted and measured measured fuel temperatures more difficult, but the conclusion temperatures conclusion reached reached during during the test programs programs was that the results were reasonably consistent were reasonably consistent with the predicted predicted values, considering considering that the non-instrumented instrumented fuel assemblies probably have as large a range range of heat transfer characteristics as the transfer characteristics instrumented elements do. Data tables from these core instrumented elements core test programs programs include include fuel temperatures temperatures as predicted predicted and measured. UWANR Safety Analysis Report Rev. 2 UWNR 436Sp.20 4-36 4-36 Sept. 2008 2008
- 16 standard First mixed core- 9 FLIP bundles and 16 bundles 55 standard bundles
- 4.5.1.3.1 This initial mixed core, Table 4-21, was operated from March 1974 Table 4-4, Figure 4-20 and Figure 4-21, 1974 I through December December 1977.
Some parameters parameters of this core were: Core Designation Designation F25-RlO F25-R10 Excess reactivity 4.05 % pP Shutdown margin 3.60 % P p Transient Transient Rod worth 1.37 % pP FP reactivity reactivity defect 1.92 % pP Peak pulse power 805MW 805 MW Prompt neutron lifetime sec 29.7E-6 sec
- 9-Bundle FLIP Core Figure 4-20 9-Bundle
- UWNR Safety Analysis Report Rev. 2 4-37 4-37 Sept. 2008
((.C)
- KW in KW ~n Pred Pred Itted.
Ictt*d Measured M~l1sur['d temperatures Temperatures'., (0 e) Core Position Core. 'Position H1eqrent F.lement TemE' 6C Temp., °c Bdle
.8d1e 41 FLIP III ,FLIP Bdle Bdle 42 FLIP 42}o'LIP Bdle Bdle 2222 STD S'l'D \) 55 NE
- 1) NE 18.1 18.1 372 372 Can't Can't Measure Measure NW NW 16.1 16.1 360 360 Cahnt-Can't Measure Measure SW SW 18.1 IB.l 372 372 Can't Cari"t Measure Measure 1.
l~ 55 NE; NK 17.4 17.4 363 363 402 402 365
}65 NW NW 16.1 16.1 350 35.0 395 395 356 350 SW 16.9 16.,9 360 360 402 402 365 365 1.7 :361
*SK rhO0 361, 361 397 397 361 K
K44 N' Ni'; 15.4 342 342 382 382 343 343 NW NW 15.4 15.4 342 342 383 383 346 346
.SW SW 16.1 16.1 350 350 384 384 348 348 SI': 17.0 17.0 361 361 380 380 343 341 1) l) 4 NK 4 ;NE 16.1 350 350 397 397 35 9 359 NW 16.1 350 353 NW 16.1 350 399 399 353 SW 17.4 363 400 363 SW 17.4 363 400 363 SE 16.2 350 398 342 SE, 16.2 350 398 342 FF 5-5 NER'.
PF 4-NE A,o', SE sit 4, NE NE 1W NW 9.3 9.3 6.7 6.7 8.0 8.0 8.8 8~8 61.3 272 272 239 239 257 257 , 266 266 290 290 238 238 260 260 280 280 233
- SW SW 6.3 238 238 233 SSE P 5.7 5*.7 223 223 216 216 FF 33 NE,NI~ 5.8 5.B 230 230 200 200 NW NW 6.8 6.8 242 2~2 222 222 SW SW .4.99 4.9 210 210 185 185 4* 6 SI;;
SL 4,'6 ,204 204 175 175 EJ NE 7.4 250 250 240 240 NW 6.6 237 269 NW .6.6
'6.*9 237 269 SW SW 6.,9 242 242 275 275 SE .7..4. 250 246 SE 7.. 4, 250 246 6.9:
DD 33 SW SW 6.9 8.3 242
.242 280 280 SE. 261 260-SF. ,8.3 261 260
- Figure Figure 4-21 4-21 Power/element Power/element and and Temperature Temperature -9 -9 FLIP FLIP Bundle Bundle Core Core UWNR UWNR Safety Safety Analysis Analysis ReportReport Rev.
Rev. 22 4-38 4-38 Sept. 2008 Sept. 2008
- Fuel temperatures in the 9-Bundle9-Bundle FLIP core are shown in Figure Figure 4-22. Instrumented Instrumented elements were located in grid position F5NE for bundle 22 (standard (standard fuel), grid position C4SW for bundle 41, and grid position E5NE for bundle 42. The instrumented element 41, element in bundle 42 was located located immediately adjacent immediately adjacent to the transienttransient rod guide tube. Although the power density was much higher in the FLIP fuel in this partial FLIP core, the fuel temperatures temperatures were reasonable.
--- ,..=-==::.:- :..:..::.... -..=.:. -:.-===-=. - -- _=:.-=-- ~-=- i
--------
-
== ,3 : -6 .:7 00. 70 :0 oo0: " :.'l~" "_":JOOO
------ --'\" . -'1
-_.. _ - - - - - - - - - -
Figure 4-22 Fuel Temperatures Temperatures vs. Power- 9 Bundle FLIP
- Power Power Defect vs. Power - 9 Bundle The power defect was strongly coefficient Bundle FLIP strongly affected by the FLIP fuel. See Figure coefficient in FLIP fuel becomes more negative during normal full power negative as temperature rises, the total power power operation is smaller, while accident response same as in standard TRIGATRIGA fuel.
Figure 4-23. Since the temperature response remains essentially power defect essentially the REACY7I7Tr z~S5::P0NER~D'E.L:7::~. - -- S-- ~?P~/~ffDA - --- -- .-
-J i I5 I I h I TM , I Z- '_7 7: *7
-7
- 1.T: - - --- -- ----
.K - :0---MO..
- - . .... -* - -- -.. :70
.. -:4W0 ','.5 . 1;,,,*
Zo.- !-- - -_÷-....... - "- -- :-:---
-- -
Figure 4-23 Power Defect vs. Power - 9 Bundle FLIP
- UWNR Safety Analysis Report Rev. 2 4-39 4-39 Sept. 2008
. Sept. 2008
- 1 4.5.1.3.2 4.5.1.3.2 Second Second mixed core- standard bundles 66 core - 15 FLIP fuel bundles and 10 standard This core, shown in Figure 4-24 and Figure 4-25, was operated operated from January 1978 until June 1979, although some initial high power operation with a core intermediate intermediate between this and the all FLIP core was done to make the initial shipment self-protecting self-protecting before the final batch of fuel was received.
received. parameters for this core were: Some parameters Core Designation Designation G25-R1O G25-RlO Excess reactivity reactivity 3.87 % Pp Shutdown Shutdown Margin 3.90 % P p Transient Transient Rod worth 1.38 % Pp FP reactivity defect 1.75 % Pp Peak pulse power 930 MW 930MW Prompt neutron lifetime 24E-6 sec
- Figure 4-24 15 Bundle FLIP core UWNR Safety Analysis Report Rev. 2 4-40 2008 Sept. 2008
*
- Temperatures (-C)
Measured Temperatures Measured (OC) KW inin Predicted Temp. Predicted Temp. Bd1e 41 Bdle 41 Bdle 42 Bd1e 42 Bdle 26 Bd1e 26 Core Position Element Element ~26 (26 or 41)*C 41)OC Center TC Center TC Center TC OS DS NE NE 17.2 17.2 363 363 CAN' CAN' T MEA MEASSURUREE NW NW 16.0 16.0 349 CANN' T CAN T MEAS MEASURE URE SW SW 17.2 17.2 363 363 CAN' 'T CAN T MEAS MEASURE URE E5 ES NE NE 16.5 16.5 355 435 NW NW 15.3 15.3 342 SW SW 15.8 15.8 347 347 361 SE 15.8 15.8 347 347 361 E4 NE NE 11.9 11.9 304 340 340 NW NW 14.6 14.6 334 355 355 SW SW 15.0 15.0 338 338 345 345 SE 13.3 13.3 318 318 322 322 04 D4 NE NE 12.5 12.5 309 309 NW NW 15, 2 15.2 340 340 SW SW 16.4 16.4 354 269 269 400 400 SE 12.6 12.6 310 310 359 E3 NE NE 9.7 9.7 277 277 237 415 415 NW NW 9.9 9.9 280 280 290 290
- 03 D3 SW SW SE NE NE NW NW SW SW SE SE n.2 11.2 10.6 10.6 11.0 11.0 10.4 10.4 10.4 10.4 11.0 11.0 294 294 288 288 293 293 285 285 285 293 247 247 294 301 250 250 FS NE NE 8.8 8.8 267 271 NW NW 8.8 8.8 267 SW SW 6.4 6.4 237 237 SE SE 6.4 6.4 237 237 237 F4 NE NE 7.5 7.5 252 252 246 NW NW 8.4 8.4 262 262 263 263 SW SW 6.1 6.1 231 220 SE 5.5 5.5 222 222 213 213 F3 NE 5.4 220 220 203 NW .6.3 6.3 236 220 220 SW 4.8 208 177 177 SE 4~5 4.5 200 200 163 163 Figure 4-25 Power/element Temperature - 15 FLIP Bundle Power/element and Temperature Bundle Core
- UWNR Safety Analysis Report Rev. 2 4-41 2008 Sept. 2008
With the .bundle -bundle core, FLIP fuel characteristicscharacteristics become more pronounced. pronounced. The standard standard fuel instrumented instrumented element in standard fuel bundle 25 was located in grid position position F5NE, and the
- measured temperatures are shown in Figure 4-26.
__ j00:: t _ I -...--.--.-- -
°C I. ~Z~T ~ -.
zoo
~7~xi ~ Iii-- - L--z-Lii-Liz---
2~izTLL::r -- z... .~. - .....------..-.-.-.--. ______
~ J
~2
- ~ _____ ____ _____ ____
________ ____ -----h-------
~~z+XhI2I~ZI774 ---------- -- ..-.
0 zz~z ~ -
* - - 17 ----- ~ ._ I rz~i
~-~Y---
___~77
- - ""
=F--
i ~! ! Figure 4-26 Standard Temperature vs. Power -15 Standard Fuel Temperature -15 Bundle FLIP FLIP Figure 4-27 shows the temperatures temperatures for the instrumented instrumented element in fuel bundle F41, F41, located located in grid position E3NE. The unusual behavior of the bottom thermocouple. thermocouple (41-B in the figure) was due to the beginnings of failure of the thermocouple thermocouple due to internal shorting.
*
.......... .... ..
---- -
-------
100 I - - - - --.--
-. *----- -- - ,=-
- -- --- ------ 1 L ] I Figure Figure 4-27 Bundle 41 Fuel Temperature Temperature vs. Power -15 -15 Bundle FLIP UWNR Safety Analysis Report Rev. 2 4-42 Sept. 2008
*
- 4~28 shows the temperature Figure 4-28 power.
500 ........ temperature for the fuel bundle F42 instrumented instrumented elementelement vs. reactor reactor 400 300 TEMP
.C T-2 00 200 100 100;---'
.J~_ -=:- n_ : -_. _~_;-. __ -: _" o---t- I - __':-.1- I t~I-I1~',i - '~"---:;:
~:I ~ -I~ -I I'I,~jl-I';~ I:-~- I-:~I~'-~- -;-I- I -J~-I --I---I-:I- I~I- I- !:1- I~:I ~I -I I-ltl il l-~-I - I1- -~'0 1EL
---~ ----~~-:=-t= r==-~ --
--
---- --
I
--------
-
- : : : : : : : - -4im--- - -5rl1r---' Ji JI: 0
~---j:.-:-:----l-
-77
~_ -l=: -- Pfl~ ----1-'------- - - - - - - :::::t===-I Figure 4~28 4-28 Bundle 42 Fuel Temperature Temperature vs. Power - 15 Bundle FLIP
- The power defect for this core is shown shown in Figure 4;.29.
lower power defect with the higher amount of power generation in a still lower 4-29. Again, the effect Pulsing behavior also showed a further reduction in the prompt neutron lifetime and this core. Pulsing thus faster periods for the same reactivity input in a pulse. effect of the FLIP fuel results generation in the FLIP portion of of
'00 Figure 4-29 Power Defect Defect vs. Power - 15 Bundle Bundle FLIP
- UWNR Safety Safety AnalysisAnalysis Report Report Rev. 2 4-43 Sept. 2008 2008
- 4.5.1.3.3 All-FLIP core7 The reactor reactor has been operated with cores containing containing only FLIP fuel since July 1979. Two variants of this core have been commonly used, differing differing only in the number of graphite reflectors used and location of irradiation facilities. The arrangement reflectors arrangement shown in Figure 4-30 with characteristics indicated below and in Figure 4-31 is the most-used variant.
the characteristics Some measured measured parameters for this core were: Core Designation Designation 123-RIO I23-R10 Excess reactivity reactivity 4.23 % P p Shutdown Margin Margin 3.80 % pP Transient Transient Rod worth 1.395% pP 1.395% FP reactivity defect 1.53 % pP Peak pulse power 950 MW 950MW Prompt neutron lifetime 23E-6 sec sec
- UWNR UWNR Safety Analysis Report Rev. 2 4 -44 4-44 Sept. 2008 2008
*
- TEMPERATURES TEMPERATURES ATAT FULL FULL POWER POWER KWKW inin Predicted Temp.
Predicted Temp. #41
#41 #42
#42 Core Position Core Position E-l~ment Element (#42)OC
(#42)°c Bot Ctr Bot ctr Top IQE. Bot Bot Ctr,
- err, TOP
~
0505 NENE 16.0 16.0 490 490 C]an Can 'tI t MMe ur e a.s ure e as NW NW '14.9 14.9 470 470 Can Can 'tI t M ea ssur Mea u re e SW SW 16.0 16.0 490 490 Can Can 'tI t ea ssur Mea M u ree E5 NE HE 15.5 15.5 480 480 365 365 382 353 382 353 NW NW 14.4 14.4 460 460 SW SW 14.6 14.6 465 465 SE SE 14.6 14.6 465 465 340 340 362 362 338 338 E4 NENE 11.9 11.9 415 415 300 300 315 315 295 NW NW 13.6 13.6 442 442 334 350 328 328 SW SW 13.7 13.7 450 450 330 330 350 326 326 SE SE 12.1 12.1 420 305 305 325 305 305 04 SW SW 15.2 15.2 475 390 405 380 492 432 407
- SE SE 15.1 151.1 470 470 426 380 343 E3 NENE 14.2 14.2 455 277 259 375 337 300 NW NW 12.2 12.2 420 275 290 270 SW SW 9.9 9.9 385 275 285 265 SE SE 10.6 10.6 395 241 224 F5 NE NE 11.6 11.6 415 415 450 400 350 SE SE 7.9 7.9 345 341 301 268 F4 NE NE 9.8 9.8 380 410 362 325 NW NW 11.0 405 442 395 352 SW SW 7.5 7.5 340 340 300 270 SESE 6.8 6.8 325 305 270 240 F3 NE NE 8.7 8.7 360 360 318 318 282 241241 NW NW 8.1 8.1 345 345 362 280 320 280 320 SW SW 5.9 5.9 300 300 220 245 280 245 280 SE SE 6.7 6.7 325 325 242 220 198198
- Figure Figure 4-31 4-31 Power/element Power/element and and Temperature Temperature -- All All FLIP FLIP Core Core UWNR UWNR SafetySafety Analysis Analysis Report Report Rev.
Rev. 22 4-45 4-45 Sept. 2008 Sept. 2008
Fuel temperatures Fuel temperatures in in the the all-FLIP all-FLIP core core are are shown shown in Figure 4-32 in Figure 4-32 and Figure 4-33. and Figure 4-33. The The bottom bottom
- thermocouple thermocouple in in fuel fuel bundle bundle 4141 had had developed developed aa short short from from one one side side of ofthe the couple couple to to ground, ground, and and thus reads thus reads well well below below the the actual actual temperature temperature inin this this graph.
graph. The The instrumented instrumented element element in in bundle bundle F41 was in grid position E3NE, while that for bundle F42 was in grid position F41 was in grid position E3NE, while that for bundle F42 was in grid position D4SW next to D4SW next to the the transient rod transient rod guide guide tube. tube.
-----t Figure 4-32 Bundle 41 Fuel TemperatureTemperature vs. Power - All FLIP FLIP 400 .,
*
!
300 0.. UJO
~ , ,-
200 100
-r; ,
Figure ~-33 4-33 Bundle Bundle 42 Fuel Temperature vs. Power Power -- All All FLIP FLIP UWNR UWNR Safety Safety Analysis Analysis Report Report Rev. Rev. 22 4-46 4-46 2008 Sept. 2008 Sept.
*
- Figure 4-34 shows the power power defect versus power for this core. In this all-FLIP core the power defect for licensed full power decreased decreased slightly from the IS-bundle 15-bundle FLIP core value. At the end of 1999, after more than 477 MWd operation, operation, the power defect at licensed full power remains at AK/K.
1.51 % ilKlK. tI 0.
- Figure 4-34 Power Defect vs. Power - All FLIP Figure 4.5.1.4 Isothermal Temperature Coefficient Isothermal Temperature Coefficient FLIP The coolant water temperature temperature in the prototype prototype was varied over wide ranges (200(20* to 60*C) 60'C) to measure the resulting reactivity reactivity change. The measurements measurements were made at power power levels of less than 10 watts. The coefficient coefficient is slightly positive with a net gain in available available reactivity 0.077%
reactivity of 0.077% over the range indicated. The average coefficient,0.00 19%/°C, is small enough that it is average coefficient,0.0019%I"C, essentially negligible negligible for normal operating operating conditions. The effect of the water gap left in the shrouds when the control blades are withdrawn was expected to increase expected increase the temperature temperature coefficient coefficient by about 20% in the UWNR, givinggiving a temperature coefficient estimated at 0.0024%/°C. 0.0024%I"C. This value was small enough to be considered considered negligible for normal operating negligible operating conditions. Values measured during startup testing were 0 and 0.0042, but with vary large uncertainty uncertainty in the values because because of other possible reactivity variations variations that might occur (bubbles, variation of rest position of control blades and the transient rod, and other extremely extremely small small variations). Considering Considering these other variations variations it is not possible to see any change in reactivity in the UWNR cores as the bulk water temperature temperature changes for either either
- the standard, mixed, or all-FLIP cores.
UWNR Safety Analysis Report Rev. 2 4-47 2008 Sept. 2008
1 4.5.1.5 Pulse Parameters Measurements Parameters Measurements were made of the various parameters reactivity up to 2.1 2.1% ~KJK for the standard-fueled
% AK/K parameters relating to pulsing operation of the prototype and of the UWNR cores. The most important of these are given below standard-fueled cores and 1.4%
below for step insertions of 1.4% AK/K ~KJK for the cores that prototype of
- contain contain FLIP fuel. The data for the prototype TRIGA TRIGA Mark III core, the UWNR UWNR standard TRIGA TRIGA core (both water and graphite graphite reflected),
reflected), and the mixed cores are indicated in the figures referenced referenced below. Period Period and Pulse Width During pulsing operation operation the "/. I
.40.o reactor is placed in a super-reactqr j :;
prompt-critical prompt-critical condition in which 1.4 .*1 the asymptotic period is related to .! : H4 I the prompt reactivity insertion '!HT{ii' i . . .1 divided into the prompt neutron J cycle time. The pulse width is I
..,..c: ' W¶"
"
inversely inversely related related to the prompt prompt o u I I; I reactivity insertion. Behavior of ~ f I the different cores and the
!
i ... j.! T
- Figure 4-indicated in Figure L
prototype is indicated 4- 1 1- Till 35 with points of inverse inverse period and FWHM FWHM shown on the same -2 .2o 1"T! 1 graph. The plots show the results ... 2 : E.
.
of plotting the reciprocal measured reciprocal of the ..
% .: '
V
- measured period versus the ; *.I" reactivity insertion. Since 5 prompt reactivity 3 gv N the period data were obtained 1 :r oscillographic recording from an oscillographic ý10 of the reactor power versus time at 1 .., " ! IT'i a portion of the pulse before fuel ~ -41 . 1
- .
temperature limiting effects temperature effects have II begun, the accuracy of the ~ Lit begun, the accuracy of the ~ . measurements is not as good as - i 1 for other parameters, and some .0 .8. ....... 2 ... ** .o for other parameters, and some scatter in the data are expected. . Promt Reactvlty, Insertion, Prompt Re8ctiyfty- 'nsertfon, ~XlK
ýaXK As can be seen, the minimum Figure 4-35 Inverse.Period and Inverse Width at Half-Max [ax period in standard standard fuel obtained obtained Figure 4-35 InversenPeriod and Inverse Width at Half-I vs. Prompt Reactivity Insertion for reactivity 2.1%
reactivity insertions of 2.1 % vs. Prompt Reactivity Insertion AK/K
~KJK is about 2.6 msec while that of the FLIP core is 3.4 msec for a 1.4% 1.4% ~KJK AK/K insertion.
As FLIP fuel replaces standard fuel in the mixed cores, the decrease in prompt prompt neutron cycle time
- results in a different different straight line plot for each core. Further, it becomes apparent on the all-FLIP UWNR Safety Analysis Report Rev. 2 4-48 Sept. 2008 2008
core core that thatthethe faster faster cycle cycle time, time, even even for for the the smaller smaller prompt promptreactivity reactivity insertion insertion allowed allowed for for the
- the mixed cores, causes mixed cores, causes the data the data for for larger larger reactivity reactivity insertions insertions to to depart depart from from aa straight straight line. line. ThisThis isis because because the the transient transient rod rod has has not not completed completed its its travel travel before before the the reactor reactor reaches reaches aa substantial substantial power power level, level, thus thus resulting resulting in in aa longer longerperiod period as as negative negative reactivity reactivity isis inserted inserted by by the the fuelfuel temperature coefficient.
temperature coefficient. Pulse Pulse Width Width 7. The width of The width power pulse the power ofthe pulse isis [z most most conveniently convenientlydescribeddescribed as as the the time interval between half-power time interval between half-power points. shown in Also shown points. Also Figure 4-35 in Figure 4-35 ". Ii . ,t.*.I..'.. 't are plots of the reciprocal of are plots of the reciprocal ofthethe : . 4 : ; :PV*i ,, il ; -... lil , measured width measured versus prompt width versus prompt reactivity reactivity insertion. insertion. Each Each of ofthethe J K ':;:- *I;i-,:- , .. : r. . cores has a different straight cores has a different straight line plot, to the increasingly again dueneutron line o 'ri :t i 'h! 1.I.L i4' 4 t ii ~i.-iL !.ll I .;. h ,ý;h
.14, '
4-........t short plot, prompt again due to the cycle time as increasingly ,,.-. . *, ! , " short prompt neutron cycle time as the the amount amount of ofFLIP FLIP fuel increases.2 2 fuel increases. , When the pulse widths (Full Width ,, : ....
~ :..... ~ , l.t;: li~iii ,
- Ih....., N'"...i-f...' I t ...
atWhen HalftheMaximum pulse widthspower) of the (Full Width t.- at Half Maximum power) of the
- 201i various cores various promptperiod prompt period as compared to are compared cores are shown in as shown to their in Figure Figure their
. ,.- :. . , :I : :,* . . .i. m..+'.. ,-L,: ..+
j* +-!..1,...4. *t m.',.. tl4.: d:lI. 441. 4-36, all of the cores conform fairly 4-36, all of the cores conform fairly -I well well toto the the same same straight straight line line n I..,--r Ii because the difference because the difference in in prompt prompt =* ,.j,,.1 . ,II.. . . .:..l 1, , neutron cycle time is not neutron cycle time is not a factor a factor - r.. ... * .l.. . in in the the relationship relationship between between these these ...
...
. . +"' .... ,.;,
- t.LJ . .
two variables as it is two variables as it is in the in the .:,
.
I i/ ................. '; ' 'j :" - "I,, F
- Lfl 1,14,+i+
4 i j' f"i+'+: reactivity-period and reactivity-period and reactivity-reactivity- <. FWHM relationships... FWHM relationships. , .. , d r~ 02 Period, Period, Milliseconds Milliseconds Figure Figure 4-364-36 FWHM FWHM vs vs Period Period
- UWNR UWNR Safety Safety Analysis Analysis ReportReport Rev. Rev. 22 4-49 4-49 Sept. 2008 Sept. 2008
- Peak Power Peak Power Peak power Peak power inin aa pulse pulse is is proportional proportional toto .:=.,
the square the square of of the prompt reactivity the prompt reactivity OiO :.4. li:i!;'i2i
. . R 01!
insertion, while insertion, while energy energy generated pulse generated in a pulse 2. . .. ' . . . ., is directly related related to the prompt prompt reactivity reactivity '.'.i I 7 insertion. Figure 4-37 shows the inter- inter- .T *, 1.,1J.* relationship relationship between between maximum maximum transient transient .I., ....i
..
power power and pulse width. Peak power power is
. ...
plotted against the plotted against square of the square of inverse inverse FWHM in order to get a straight line plot. FWHM 1200,<.... The standard standard and and FLIP fueled cores show 4 11; differences differences due to the shorter prompt .l" neutron cycle time. t Figure Figure 4-38 4-38 shows the relationship relationship Inverse Half-Width Squard (1)! Millisecood1.e 103 between between initial initial period period and peak power and seems to be fit fairly well by a single Figure 4-37 Peak Power Power vs Inverse Inverse Half-width Half-width
- straight line. Note that the prompt prompt Squared Squared reactivity insertion is limited to 1.4 % %
AK/K for the FLIP cores. ilKlK For a given configuration, the core configuration, given core peak the peak 0' 1 power, integral integral power in the prompt burst, and width of the pulse are determined determined by .. iq--- . the reactivity reactivity insertion made. It can be Mi I; seen from the plots that the peak power is -* .,* I controllable over a rather wide range since ! J'1 this parameter is very nearly proportional ! . T, (AK/K- - 0.7%)2. to (ilKlK 0.7%)2. Pulse width and It! J. R;I integral powers, on the other hand, are approximately linear functions of of reactivity insertions above prompt critical .400 ........ so that their range is more limited. 00o ... ........................
.I..o
. o00 N.ewtor Perfod,
_.ector Miiiise-ids perlid, Mllilleconds Figure 4-38 Peak Power vs. vs. Reactor Period 0 UWNR Safety Analysis Report Rev. 2 4-50 Sept. 2008 2008
- 4.5.2 Operating Limits Operating
- For previous previous operation of the reactor at 1 MW, the reactivity above clean cold critical. The reactivity is allocated reactivity has been less than 4.9 % ilKlK approximately as indicated below:
allocated approximately Power Coefficient Coefficient 1.75% 1.75% AK/K AK/K t.KlK Xenon Poisoning 1.75% 1.75% LK/K t.KlK Control & & Flux Balancing Balancing 1.40% 1.40% AK/K t.KlK No specific amount of negative negative reactivity available from control element action is specified, since the requirement on minimum shutdown shutdown margin assures assures safe shutdown shutdown from any operating operating condition. The minimum shutdown margin with the most reactive reactive control element element and any un-scrammable un-scrammable control elements full out will not be less than 0.2% ilKlK. AK/K. Shutdown margin is verified verified by calibrated control element element positions and by rod-drop rod-drop measurements. measurements. The limitations on cores cores containing FLIP fuel will maintain power power density to levels capable of capable of natural convection convection cooling during power operation up to 1.5 MW power. This limitation will also assure that power power density in any fuel element will be below below that at which loss of reactor reactor coolant coolant will result in fuel damage.
- In addition, the maximum maximum reactivity reactivity for an experiment experiment is limited to 1.4 % ilKlK AK/K All in-pool experiments experiments will be constrained constrained at least as well as the fuel bundles. In-core In-core experiments experiments are designed so they are constrained constrained by the grid or grid box structure, although part of their support may be from other pool structure.
Should an experiment experiment having the maximum reactivity worth allowed for all experiments experiments (1.4% (1.4% AK/K) fail, the resulting step change in reactivity worth would be less than that deliberately ilKlK) inserted during pulsing operation. Should the beam ports and pneumatic tube flood while the reactor is operatingoperating at full power, a step reactivity reactivity addition of 0.07% 0.07% ilKlK AK/K would result. This reactivity change change is so small that it would not cause any disruption of normal operation. If a gross departure procedure were to be made and a fuel element bundle were added to the departure from procedure outside of the core while operating operating at full power, the maximum reactivity reactivity that would result would would be about 0.7% 0.7% AK//K. ilKlIK. This is a reactivity reactivity smaller smaller than that routinely inserted during pulsing operation. operation. Despite the built-in safeguards and inherent safety of the reactor and its fuel, great attention is paid to proper supervision supervision of operation operation and adherence to procedures procedures approved by competent competent authority. It is the policy of the University of Wisconsin Wisconsin that standard standard operating operating procedures procedures are carefully carefully prepared prepared and reviewed, strictly followed, and kept current. Likewise, competent competent
- UWNR UWNR Safety Analysis Report Rev. 2 4-51 2008 Sept. 2008
- supervision supervision assures that operation operation is kept within the limits set by licenses, technical specifications, existing procedures, specifications, procedures, and general good practice.
4.6 Thermal-Hydraulic Design Thermal-Hydraulic General Atomics has done extensive extensive thermal-hydraulic computations over the years, including thermal-hydraulic computations one study specifically specifically directed at the use of four-element bundles ofTRIGA of TRIGA fuel as a replacement fuel for reactors which were originally replacement originally fueled with flat-plate fuel elements elements and 8 operated at power levels up to 2,000 kW with natural convection 8 convection cooling *. The Puerto Rico Nuclear Center TRIGA-FLIP TRIGA-FLIP reactor used 4-element 4-element fuel bundles and operated operated at power levels much higher than the 1,000 kW steady-state power power level ofUWNR of UWNR with natural convection convection cooling 99.* The thermal and hydraulic hydraulic design for operation of the Torrey Pines Thermionic Reactor, section 3.3 of the reference, reference, describes the core parameters parameters for a very lO core'°. very similar core
- The conclusions references were that four element bundles ofTRIGA conclusions of these references of TRIGA fuel could be used for power levels up to 2000 kW with naturalnatural convection cooling. The University Wisconsin University of Wisconsin Nuclear Reactor Reactor has been operated operated with natural convection convection cooling at steady-state power levels levels up to 1,000 kW for many years with no cooling problems problems and no fuel damage. Many other TRIGA TRIGA reactors have been operated operated at power levels up to 1.5 MW with natural convection convection cooling and no cooling problems.
4.7 References 1. 1. 2. Reactor Shielding Design Manual, Theodore Company, 1956, 1956, page 178. No. 4, Report Memo No.4, Theodore Rockwell Rockwell III, Editor, McGraw McGraw Hill Book Report on Refueling the University of Wisconsin Nuclear Reactor, R. J. Cashwell, Nuclear Nuclear Engineering Departmen~, Department, University University of Wisconsin, Wisconsin, March 1968. Book 1968.
*
- 3. GA-9064, Safety Analysis Report for the Torrey Pines TRIGA TRIGA Mark III Reactor, Section Section 3.2 and Figure 3-16, General Atomics, Jan. 5, 1970.
- 4. Same as 2.
5.
- 5. Core Test Program UWNRUWNR Mixed TRIGA-FLIP TRIGA-FLIP Core (9 FLIP Bundles),R. J. Cashwell, Nuclear Nuclear Engineering Engineering Department, University of Wisconsin, July 1974. 1974.
- 6. Core Test Program UWNR Mixed TRIGA-FLIP Core (15 (15 FLIP Bundles),R. J. Cashwell, Cashwell, Nuclear Engineering Nuclear Engineering Department, University of Wisconsin, February 1978.
- 7. Core Test Program All FLIP Core) ,R. J. Cashwell, Cashwell, Nuclear Nuclear Engineering Engineering Department, University of Wisconsin, January 1980.
8.
- 8. Steady State Thermal Analysis for the Proposed Proposed Use ofTRIGA of TRIGA Fuel Elements in MTR MTR
- Reactors, GA-5708, General Atomics, Reactors, GA-5708, 1965.
Atomics, 1965. UWNR UWNR Safety Analysis Analysis Report Rev. 2 4-52 2008 Sept. 2008
99.. Safeguards Summary Report TRIGA-FLIP Reactor at the Puerto Rico Nuclear Report for the TRlGA-FLIP
- 10.
Center, PRNC 123, Puerto Rico Nuclear Center, November 11, Safety Analysis Report for the Torrey Pines TRIGA General Atomics, January 5, 1970. 1970 . 11, 1969. Gulf TRlGA Mark III Reactor, GA-9064, Gulf
- UWNR Safety Analysis Report Rev. 2 4-53 Sept. 2008 2008
- This page is intentionally intentionally left blank.
- UWNR UWNR Safety Safety Analysis Analysis Report Report Rev. 2 4-54 4-54 Sept. 2008 2008
- 55 REACTOR COOLANT SYSTEMS
- COOLANT SYSTEMS 5.1 5.1 Summary Description Summary Description schematically in Figure 5-1.
The pool water is cooled by the system shown schematically 5-1. The design basis ofof the reactor coolant system is to dissipate 1.0 MW with primary temperatures temperatures approximately approximately 80 'Fof and prevent prevent the inadvertent loss of pool water. The system, however, performs no safety function. The system consists of three loops; the closed-loop primary coolant system, the closed-loop closed-loop primary closed-loop intermediate coolant system and the closed-loop intermediate closed-loop campus chilled water system. Heat from the primary coolant coolant system is transferred to the intermediate coolant system through through the primary heat exchanger. Heat from the intermediate intermediate coolant coolant system is then rejected to the campus chilled water system through the intermediate heat exchanger. The system is designed to maintain a pressure gradient towards the pool in orderorder to prevent the inadvertent loss of pool water. 5.2 Primary Coolant System The primary coolant coolant system is composed of a pump, isolation isolation valves and various devices used to extract flow rate, temperatures temperatures and pressures. Stainless Stainless steel components components and piping are used in the system in order to maintain maintain primary water quality more easily. The primary system continuously continuously circulates circulates pool water through the primary heat exchanger. The intake and outlet
- diffusers include siphon breaker holes to preclude preclude draining more than 1 foot of water even in the case of a pipe rupture. This will maintain at least 19 feet of water above the active core.
5.3 Intermediate Intermediate Coolant System The intermediate intermediate coolant coolant system consists of a pump, isolation valves and various devices used to extract temperatures temperatures and pressures. Stainless steel components components and piping are used in the system system to maintain reactor grade quality water in the intermediate coolantcoolant system. Circulation in this maintained by the intermediate system is maintained intermediate pump, discharging discharging through the intermediate intermediate heat exchanger, where it will reject heat to the campus chilled water system. The cold water will then circulate through the primary heat exchanger exchanger to cool the primary primary water and return to the pump suction. The intermediate intermediate coolant system is equipped equipped with a pressurized pressurized expansion tank and a make-up water system. The expansion tank will accommodate accommodate volumetric volumetric changes changes in the intermediate system process fluid and maintain the intermediate intermediate system pressure above the primary coolant system pressure under both static and operational conditions. By maintaining the intermediate system pressure pressure higher than the primary system, should a leak occur, it would result in intermediate water entering the primary intermediate primary system thereby protecting protecting against the inadvertent inadvertent loss of of pool water. A pressure pressure sensor provides indication indication at the control console console and an interlock interlock prevents prevents starting the primary primary pump unless the intermediate intermediate loop pump is running. running .
- UWNR Safety Analysis Report Rev. 1I UWNR Safety 5-1 5-1 August 2004
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- 0
- Should leakage occur, it could be detected in three three ways. First, the intermediate intermediate loop will be unable to maintain pressure and a low pressure annunciator will alarm at the reactor control console. Second, the pool level float switch will be actuated by high pool water level should as much as 150150 gallons of intermediate intermediate water enter the pool system. Finally, if the integrity ifthe of the integrity ofthe intermediate water heat exchanger is also compromised, intermediate compromised, an influx of degraded degraded quality intermediate water will increase conductivity in the pool water.
intermediate 5.4 Campus Chilled Water SystemSystem The campus chilled chilled water system consists of carbon steel piping, pump, isolation valves and various devices devices used to extract temperatures and pressures. Circulation Circulation in this system is maintained maintained by the chilled water pump, taking a suction on the main campus campus chilled water system. The pump discharges discharges through a filter and into the intermediate heat exchanger, exch~nger, where it cools the intermediate loop water and returns to the main campus chilled water system. The campus chilled water loop is maintained at a higher pressure than the intermediate intermediate system. This pressure gradient gradient will insure that in the extremely unlikely event of leaks in both the primary coolant coolant system and intermediate coolant system heat exchangers exchangers and loss of intermediate system pressure that inadvertent inadvertent loss of pool water will be physically physically impossible. 5.5 5.5 Primary Coolant Cleanup Cleanup System
- Water connections connections through the biological shield are shown in Figure system is shown schematically schematically in Figure the pump, through the demineralizer, Figure 5-3.
Figure 5-2. The pool clean-up 5-3. Water is circulated clean-up circulated from the pool surface, through demineralizer, and then into the pool under the core box and coolant header. The pump maintains about 18 gallons/minute flow through the demineralizer. The demineralizer is a mixed-bed type with provisions demineralizer provisions for regeneration of resins or discharge discharge of spent resin and loading with new resin. A water softener supplies softened water for regeneration regeneration ofof the demineralizer. Flow from the demineralizer demineralizer to the pool is through valve 10102, 10102, check valve 22 which prevents back flow, and valve 719 into the 8 inch pipe loop and into the bottom ofthe of the grid box. The 8 inch line is equipped with a siphon breaker at the top ofthe of the pool so that rupture ofthe of the line at the demineralizer outlet or of the 8 inch line outside the shield cannot demineralizer cannot drain the pool to a level that will uncover the core. A second 8 inch line is flanged off on both ends. The 8 inch lines were originally originally installed to allow a forced-convection forced-convection cooling mode, but the lines are used only as indicated above. A two inch line whose rupture could have caused loss of pool water has been permanently permanently plugged inside the concrete concrete shield and is presently sealed off outside the shield. A pool drain drain line and valve have been eliminated. There There are no valves in the system that, if opened, can drain drain the pool.
- UWNR Safety Safety Analysis Report Report Rev. 2 5-3 5-3 2008 Sept. 2008
Should valve number 5 (shown in Figure 5-3) 5-3) be left open upon placing the system in its normal
- operating condition, as much as 400 gallons of pool water could be pumped to the holdup tank. tanle No further loss of water would then occur, since checkcheck valve 22 will prevent prevent reverse flow from the 8 inch pipe loop to the demineralizer demineralizer and the siphon breaker breaker at the top of the loop will prevent prevent additional water loss.
demineralizer regeneration Wastes from demineralizer regeneration and waste poured down the reactor reactor laboratory floor drain or radioactive radioactive sink are collected in the waste system holdup tank. The waste system consists of a 2000 gallon holdup tank, pump, and filter, and is shown schematically schematically in Figure 5-4. The holdup tank is periodically periodically sampled, analyzed, and then pumped pumped out into the sanitary sanitary sewer through through 0.5.micron OSmicron filters to preclude preclude any particulate particulate activity from being discharged. All operations operations involving the cleanup cleanup system are performed performed by written checklist-type checklist-type procedures procedures designed designed to prevent draining of the pool. 5.6 Primary Primary Coolant Makeup Makeup Water System The pool makeup makeup water system is shown schematically schematically in Figure Figure 5-5. Normally, makeup makeup water water is supplied by the still. The still delivers water to a system of storage tanks from which which it is pumped pumped (by the pool recirculating recirculating pump) into the pool to maintain pool water level. Although Although distilled water is normally used for makeup, alternate flow paths allow softened or city water to be fed through the demineralizer demineralizer into the pool. In either case, impurities in make-up water are
- reduced to less than 1 ppm before going into the pool.
All operations involving the makeup system are performed by written checklist-type procedures All operations involving the makeup system are performed by wrItten checklist-type procedures designed to prevent draining of the pool. 5.7 Nitrogen-16 Control System Nitrogen-16 Nitrogen-16 suppression is accomplished accomplished by aajet-type jet-type diffuser system. This system pumps about 80 gallons of water per minute from near the pool surface through a single nozzle having a 0.75 inch wide, 6.5 inch long opening. The nozzle is located 5.5 feet above and 0.5 feet east of the core, with the diffusing stream directed downward downward at a 45 degree angle toward the west end of of the pool. The pump for the N-16 N -16 suppression system is located located on the outside east face of the reactor's reactor's concrete pool shield structure, about 8 feet below the pool surface. The system is constructed constructed with siphon breaker holes which preclude draining draining more than one foot of water from the pool in the event of a pipe rupture. 5.8 5.8 Auxiliary Auxiliary Systems Systems Using Primary CoolantCoolant There There are none. UWNR UWNR Safety Safety Analysis Analysis Report Report Rev. 2 5-4 2008 Sept. 2008
*
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- Figure Figure 5-25-2 Pool Pool WaterWater Systems Systems UWNR UWNR SafetySafety Analysis Analysis Report Report Rev. Rev. 22 5-5 5-5 Sept. 2008 Sept. 2008 System Water
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Figure S-S 5-5 Water Makeup System Schematic Schematic 5.9 S.9 References References There are none. UWNR UWNR Safety Analysis Report Rev. 2 5-8 5-8 Sept. 2008
- 66 ENGINEERED SAFETY SAFETY FEATURES FEATURES
- ENGINEERED 6.1 6.1 Summary Description Summary Description Engineered safety features are not required Engineered required for this reactor due to low operating operating power and good fission product retention in the fuel. A product retention A confinement controlled ventilation confinement with a controlled ventilation system is provided, however, to reduce the consequence provided, consequence of fission product releaserelease from fuel or experiment experiment malfunctions malfunctions to even lower levels 6.2 6.2 Detailed Detailed Descriptions Descriptions 6.2.1 6.2.1 Confinement Confinement The The Reactor Laboratory is a 43 by Reactor Laboratory by 7070 foot room of conventional conventional construction construction within the Mechanical Engineering Mechanical Engineering Building ,with a ceiling height approximately 36 height of approximately 36 feet in most ofof the room. The portion of the ceiling above the consoleconsole area is at a height of 22 feet. Figure of22 Figure 6-1 6-1 through Figure Figure 6-6 6-6 show show the outlines of the room and location of major reactor components.
components. concrete laid on the ground. The floor of the room is concrete ground. The The walls are concrete and brick. brick. The 1-1/2 inch steel deck with 2 inches ceiling is a 1-1/2 4-ply, built-up surface. inches of rigid insulation and a 4-ply, console area is located The console located in the southwest corner comer of the Reactor Reactor Laboratory. Laboratory. It is separated separated on
- the north and east sides from the laboratory noise originating Laboratory. The non-shared Laboratory.
exhaust non-shared use Reactor proper by laboratory proper originating from the cooling system and other pumps the east and one on the north of the console Reactor Laboratory exhaust air to the console area. The console by wire reinforced reinforced glass Laboratory ventilation glass provided provided to reduce pumps and equipment. Two doors, one console area open into the remainder remainder of the Reactor ventilation system therefore within the console is therefore system provides Reactor reduce one on provides both supply the confinement on supply and confinement system of the Reactor Laboratory Reactor Laboratory The Reactor Laboratory has no exterior windows, Reactor Laboratory windows, but the control control room does have a single borrowed-light borrowed-light window on the south south side visitor's center, room 1101, side looking into the visitor's 1101, which which has two large windows facing the parking parking lot and stadium. stadium. The control room window does not open. open. There are three single doors; one opening opening at ground level on the control control room south south wall into the visitor's center center (which connects connects to the Mechanical Mechanical Engineering Engineering lobby through a locked door), opening at ground one opening ground level on the east wall into a locked vestibule (which connects connects to a public hallway), and one opening opening at basement level on the west wall into the Reactor Laboratory Reactor Laboratory auxiliary auxiliary support space. One double door opens at basement basement level on the east wall into the Reactor support space. All Laboratory auxiliary support Reactor Laboratory All doors have narrow viewing viewing windows, are interior doors which which do not open to public, un-restricted un-restricted areas, areas, and are fire doors and are not weather-stripped. special seals weather-stripped. No special seals are provided for lines that penetrate penetrate the walls. All doors are walls. All normally normally closed and locked for security security and air flow control considerations considerations..
- UWNR Safety Analysis Report Rev. 2 UWNR 6-1 61Sp.20 6-1 2008 Sept. 2008
The Reactor Laboratory Laboratory auxiliary north, and east basement basement level rooms having concrete pneumatic tube equipment level (See Figure 6-1). concrete walls on the order of 8 equipment (B (B 1135D), air activity monitor (Bl135D), monitor equipment (B surround s the Reactor Laboratory on the west, auxiliary support space surrounds 6-1). This auxiliary support support space contains contains small 8 inches thick. The small rooms house the 1135C) and dispatch station (B (Bl135C) (Bl135E), (Bl135B), 1135B), a sample preparation room 1135E), and both general storage (B (B1215D) 1215D) and
- radioactive radioactive storage areas (BI135A (Bl135A andB1215C).
and.B1215C). A small small hot cell is located on the east end (B1215C), (B 1215C), an instrumentation instrumentation shop is located located in the north-west corner, comer, while the west end of the room is a counting laboratory laboratory with HPGe, HPGe, Nal, NaI, and proportional swipe counters. The east end is used for teaching nuclear engineering laboratory nuclear engineering laboratory classes (NE 427 and 428). The ground-floor of the Mechanical Engineering ground-floor level ofthe Engineering Building to the east of the Nuclear Reactor Laboratory houses an office area for the Reactor Laboratory. The Reactor Reactor Laboratory is a restricted area. All doors are kept locked locked at all times except when when authorized personnel authorized personnel are in the room. Keys are issued to a small number authorized number of authorized personnel. 6.2.2 Containment Containment No containment is needed or provided.
- 6.2.3 Emergency Emergency Core Cooling System emergency core cooling system is required No emergency required due to the low operating power.
6.3 6.3 References References There There are no references for this chapter. UWNR Safety Analysis Report Rev. 2 6-2 6-2 2008 .' Sept. 2008
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- I I
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- UWNR UWNR Safety Analysis Report Rev. 2 6-3 6-3 Sept. 2008
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- I Figure 6-2 6-2 Reactor Reactor Laboratory First Floor Plan UWNR UWNR Safety Analysis Report Rev. 2 6-4 6-4 Sept. 2008
*
- 6-3 Figure 6-3 Reactor Confinement Cross Section Through Core Centerline, Facing South Reactor Confinement
- UWNR Safety Anslysis Report Rev. 2 6-5 6-5 Sept. 2008 .
- I I.
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I I I I I I I I Figure 6-4 Reactor Confinement Cross Section Through Core Centerline, Reactor Confinement Centerline, Facing North I UWNR Safety Anslysis Report Rev. 2 6-6 6-6 Sept. 2008
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- 6-5 Figure 6-5 Reactor Confinement Cross Section Through Core Centerline,
- Facing East UWNR UWNR Safety Anslysis Report Report Rev. 2 6-7 6-7 2008 Sept. 2008
*
- melmellt Cross SectIOn Through Core Centerline, Facing West T
UWNR U WNK Safety Analysis Analysis Report Keport Rev. Key. 2 6-8 6-8 Sept. 2008 2008
- INSTRUMENTATION AND AND CONTROL SYSTEMS
- 7 INSTRUMENTATION CONTROL SYSTEMS 7.1 7.1 Summary Description Summary The reactor reactor operates in three standard standard modes:
Mode 1 Manual or automatic operation operation at power levels up to 1,000 KW. Mode 2 Square-wave Square-wave operation (reactivity insertions to reach a desired steady state power level essentially instantaneously) instantaneously) at power power levels between between 100 100 and 1,000 KW. Mode 3 Pulsed operation produced by rapid transient rod withdrawal withdrawal that results in a step insertion of reactivity up to the reactivity reactivity limit established established in the Technical Technical Specifications. Specifications. A selector switch is provided to select manual, automatic, square-wave, or pulsing modes of automatic, square-wave, of operation. Operation is from a console displaying all pertinent console displaying pertinent reactor operation operation conditions. Instrumentation Instrumentation is entirely analog, except for a digital chart recorder, digital fuel temperature temperature indication, and a computer based pulse recorder recorder used to display the power trace during pulsing operation. 7.2 7.2 Design of Instrumentation Instrumentation and Control Systems radiation-based power instrumentation, with significant overlap, are provided to Four ranges of radiation-based
- cover the operating temperature operating range from source level to the maximum permitted pulse power. Fuel temperature is also measured process variables variables are measured and used by the reactor measured, but not reactor protection protection system. In addition, other used in the reactor instrumentation and control system for the UWNR.
the instrumentation other reactor protection system. Figure 7-1 shows 7.2.1 Design Criteria Criteria The instrumentation instrumentation and control system provides the following functions:
** Provides the operator with information on the status of the reactor reactor 0* Provides Provides the means for insertion and withdrawal withdrawal of control elements ** Provides Provides for automatic control of reactor power level "* Provides the means for detecting over-power or fuel over-temperature detecting over-power over-temperature and automatically scram the control elements elements to terminate the condition condition "* Provides auxiliary trip functions based on possible loss of operability of the channels channels providing providing the overpower overpower protection protection *" Provides a record record of operation operation and radioactivity discharged from the stack radioactivity discharged
- UWNR UWNR Safety Analysis Report Report Rev. 2 7-1 7-1 2008 Sept. 2008
c UWNR Safety Analysis Report Rev. 2
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- 7.2.2 Design-Basis Requirements Design-Basis Requirements The primary design basis for TRIGA reactor safety is the safety limit on fuel temperature. A trip on high fuel temperature temperature is set to assure that the fuel temperature temperature will not be exceeded. Since fuel temperature measurement measurement includes includes time lag due to thermocouple thermocouple response response time, a reactor reactor trip based on reactor power level as measured measured by a neutron sensing system is provided.
7.2.3 System Description Description 7.2.3.1 Start-up Start-up Channel As shown in Figure 7-1 7-1,, the sensing element element for this channel is a fission counter with a drive which can be positioned positioned by the console operator. The counter has a range from 2 nv to 106 106 nv. Since the counter is moveable, its effective range is thus from about 2 micro-watts micro-watts to 22 MW. The pulses from the startup startup counter are amplified and converted converted to a logarithmic count-rate logarithmic count-rate displayed on a meter and recorded. The amplified displayed amplified pulses may also be sent to a scaler scaler that is used for subcritical measurements. measurements. The amplifier includes a normally open relay which allows control control element element withdrawal if the count rate is greater withdrawal only ifthe greater than 2 counts/second. Another Another normally closed relay provides protection to the fission counter by preventing preventing insertion of the fission counter drive when the count rate is too high. The start-up channel,channel, in the full in position, safety channel overlaps the low end of the safety overlaps channel range instruments.
- 7.2.3.2 Log N - Period Channel This channel monitors the power level of the reactor over the range from 0.1 watt to full power.
The Log N - period amplifier detects the signal from a compensated amplifies the signal signal to provide provide a 7-decade 7-decade logarithmic compensated ionization ionization chamber chamber and logarithmic display proportional to power level. The amplifier also extracts extracts period period (startup (startup rate) information. The Log N signal is recorded and operates a normally open relay used in pulse and square square wave modes modes to prevent firing the transient rod when above above 1 kW. The period signal is recorded, displayed on a meter on the console, and fed to the automatic control channel when the mode switch switch is in AUTO AUTO mode. 7.2.3.3 7.2.3.3 Pulse Power Power Channel Channel Current Current from a gamma ionization chamberchamber (or an uncompensated uncompensated neutron ionization ionization chamber) is fed to a digital data acquisition "PULSE" mode, a signal concurrent acquisition channel. In "PULSE" concurrent with firing the transient transient rod causes data to be recorded and displayed on the console computer. Information Information on on peak peak power and integrated power in the pulse is automatically automatically computed and displayed. Peak fuel temperature temperature after the pulse is also recorded and displayed on the console console computer.
- UWNR Safety Analysis Report Rev. 2 7-3 7-3 Sept. 2008 2008
I 7.2.3.4 Safety from each Safety Channels Two safety channels channels monitor reactor power level from about 0.1 watt to full power. The signal each channel channel originates originates in a compensated ionization chamber. The chamber signal is fed compensated ionization into a solid state picoammeter. The picoammeter includes normally open relay contacts which which
- open on an overpower condition to cause a reactor scram. Should either overpower condition either or both safety channel channel scram signals be present, the reactor shuts down. The scram The power level scram scram trip point is set to 1.25 1.25 times the maximum operating maximum operating level. The safety safety channels provide an additional interlock interlock signal to pulse and square wave logic to prevent firing the transient rod unless the picoammeters picoammeters are on the full-power range.
full-power 7.2.3.5 Temperature Temperature Measurements Measurements Fuel element internal temperature is indicated at the console. It causes an alarm and scram at the limiting safety system setting. The temperature temperature of the bulk pool water is measured at the core inlet by a thermocouple. This temperature indicated on the console temperature is indicated console recorder and causes an alarm and a scram on high temperature. Primary intermediate cooling, and campus chilled water systems Primary cooling, intermediate systems inlet and outlet outlet
- temperatures, and demineralizer demineralizer inlet temperature temperature are indicated on the console recorder. An alarm alarm on this recorder indicates temperature at any of these points.
indicates excessive temperature 7.2.4 System System Performance Analysis Performance Analysis The instrumentation instrumentation and control systems have been in routine operation for the almost 50 years of operating operating history. All of the measuring measuring instruments instruments have been replaced replaced over the years with instruments instruments incorporating incorporating advances advances in electronics electronics but meeting exceeding the original design meeting or exceeding criteria. The switch to entirely solid-state solid-state electronics improvement in .. electronics has resulted in marked improvement stability stability and reliability. Limiting Limiting safety system settings, limiting limiting conditions conditions for operation, operation, surveillance surveillance requirements, and action statements concerning the control control and instrumentation instrumentation systems are detailed in the proposed proposed Technical Technical Specifications Specifications in Chapter 14. 7.2.5 Conclusion Conclusion Operation during the term of the license license has shown the instrumentation instrumentation and control system to be capable of performing all intended functions with excellent excellent stability and reliability. The system system* is expected to continue to perform the intended functions. UWNR Safety Analysis Report Rev. 2 7-4 7-4 2008 Sept. 2008
- by the safety
- The The safety safety limits of fuel temperature temperature and reactor reactor power power are adequately adequately protected protected by safety channels channels and the fuel temperature temperature channels, each each of which will cause cause a reactor shutdown shutdown if the safety system settings are exceeded.
limiting safety Additional components exceeded. Additional components which cause cause trips on loss conditions necessary of conditions necessary for continued continued power power operation operation (loss of high voltage voltage to detectors, detectors, loss ofof water from the pool, pool, pool water temperature above the temperature water temperature temperature used in calculation calculation of event event consequences, loss of electrical consequences, electrical power) power) provide assurance that provide assurance that the primary primary protection protection equipment equipment will operate as planned. planned. 7.3 7.3 Reactor Reactor Control Control System System 7.3.1 7.3.1 Mode Mode Switch The mode switch switch and associated logic circuits provide the following capabilities and operating following capabilities operating restrictions different positions. The switch is a cam operated restrictions for the different operated switch which is rotatedrotated through consecutive positions. From" through consecutive From " MANUAL", MANUAL", rotating clockwise clockwise places the switch in "AUTO" mode; rotating counter-clockwise "AUTO" "SQUARE WAVE" counter-clockwise places the switch first in "SQUARE WAVE" mode, "PULSE" mode. Returning then in "PULSE" "PULSE" to "MANUAL" Returning from "PULSE" "MANUAL" requiresrequires going through the "SQUARE WAVE" "SQUARE WAVE" position. Table 7-1 details the conditions Table 7-1 conditions and restrictions restrictions invoked invoked in the different mode switch different switch positions. 7.3.2 7.3.2 Manual Operation Manual Operation
- For manual operation automatic operation the control level. At this level control elements level the reactor elements are slowly automatic control. The automatic control channel maintains regulating blade, transient rod, or #2 control blade. Figure control system.
withdrawn to obtain slowly withdrawn reactor may continue to be operated maintains power Figure 7-1 7-1 shows obtain the desired operated manually or it may be switched level by power level by servo power desired power switched to servo control shows a block diagram of the of the control ofthe 7.3.3 7.3.3 Square Operation Square Wave Operation This mode is provided for those applications applications which which require that the power levellevel be brought rapidly to some high level,level, held there for a period period of time, and then reduced rapidly producing then reduced producing a square wave of power. In the square square wave mode the reactor reactor is brought brought to a level 1000 watts in the manual level of less than 1000 mode. The mode switch switch is then changed changed to the square wave position. Changing Changing of the mode mode switch to the "SQUARE "SQUARE WAVE" WAVE" position removes an interlock that prevents prevents application application of airair to the transient rod unless the transient rod is in the full "IN" "IN" position. A A preadjusted preadjusted step change iis's then made to bring the reactor to preset reactivity change reactivity preset power levels between between 100 100 and 1000 1000 kW. The reactivity reactivity step change is made with the transient transient rod. The automatic automatic control system control system inserts inserts additional additional reactivity required to maintain the preset preset power level as the fuel fuiel heats up. The The operator operator must manually augment reactivity inserted augment the reactivity by the servo inserted by servo because because the transient transient rod does not have sufficient sufficient worth to overcome overcome the power defect at high power power levels. The linear
- UWNR Safety UWNR Safety Analysis Report Report Rev. 2 7-5 75Sp.20 7-5 Sept. 2008 2008
power level scram is maintained maintained at 1.25 P max. and an interlock prevents prevents initiation of this mode
- if the range switch is not on the full power range setting.
7.3.4 Pulsing Operation Operation The reactor reactor is brought to a power level of less than 1000 watts in steady state mode. The mode switch is then changed changed to pulsing mode. When the switch is in pulsing mode the normal neutron channels channels are disconnected disconnected and a high level pulsing chamber is connected to read out the peak power of the pulse on the console computer. Changing of the mode switch to the "PULSE" "PULSE" position removes an interlock interlock that prevents prevents application application of air to the transient rod unless the transient rod is in the full "IN" position. Fuel temperature temperature is recorded recorded during pulsing operation. The pulse channels channels are also indicated indicated on Figure 7-1 7-1. .
- 0 UWNR U N~ Safety Sa L:.*y Analysis nali~l/5:
ys s Report
]3I Rev.
£pitRv. 2
/2 7-6
/-U Sept.
Sept. 2008
/-VVO
- Table 7-1
- 7-1 Switch Functions Mode Switch Mode Switch Switch Conditions/restrictions Conditions/restrictions Position Position Manual MANUAL Transient rod may be fired only if scram is reset and transient rod rod drive is at the "IN" "IN" position.
Automatic AUTO Same as Manual, Manual, except automatic control system can control control power, subject to period limit Square-Square- Transient rod can be fired from other than full in position only if SQUARE Transient Wave WAVE WAVE scram is reset, both safety channels channels are on top range, and power power level does not exceed 1 kW as indicated indicated by the LogN channel. channel. The period channel in the LogN amplifier is defeated (restored after a short time delay when the switch is returned to "MANUAL" "MANUAL" position). Automatic control system can control power if actual power power is within +/-5% +5% of scheduled power. Pulse PULSE Period channel remains defeated. Prohibits control blade withdrawal. If the scram is reset, both safety channels channels are on top range, and power level does not exceed 1 kW as indicated' by the LogN channel; (a) Transient rod can be fired from other than full in position, (b) High voltage is removed from the fission counter, safety channel CICs , and the LogN CIC. channel CIC .
- (c)
(c) (d) Signals from all CICs are directed to ground rather than to instrument input. instrument The transient transient rod drops automatically 15 seconds or less after it is fired (e) The pulse power level channel channel is sent a signal causing it to record the pulse power trace. When returning to the "SQUARE/WAVE" "SQUARE/WAVE" position, high voltage is restored to the detectors immediately immediately and the signals to the neutron neutron measuring measuring instruments are restored after a short time delay (to prevent damage to instrument inputs from the transients resulting 1 from high voltage restoration). restoration) .
- UWNR UWNR Safety Analysis Report Rev. 2 7-7 7-7 2008 Sept. 2008
1 7.3.5 7.3.5 blade. Control Control Element The following conditions Operation Element Operation There are five control control elements; shim-safety blades, a transient elements; three shim-safety blade.. The shim-safety bladesblades and the transient control transient control conditions must be met before any control element drive control rod, and a regulating control rod have scram capability. capability. drive can be withdrawn withdrawn
*
(raised), either (raised), manually or by either manually by the automatic automatic control system: 1.
- 1. No scram conditions present scram conditions present and scram relays reset;
- 2. startup channel Count-rate on startup channel greater thanthan 2 counts counts per second; 3.
- 3. Fission Counter not in motion;
- 4. Console key switch set to "ON" "ON" position; position; There are no interlocks interlocks or permissives which restrict element restrict insertion (lowering) of control element drives. Insertion accomplished by Insertion is accomplished placing the individual by placing individual momentary-contact momentary-contact control control switches switches "IN" position, in the "IN" by a maintained-contact position, or by "RUNDOWN" switch maintained-contact "RUNDOWN" switch which which inserts all control control element drives to the "IN""IN" limit. The three shim/safety shim/safety blade drives also automatically automatically run to the "IN" limit when "IN" when a SCRAM occurred.
SCRAM has occurred. I7.3.6
- 7.3.6 Safety Blade Control The three safety safety blades are controlled by are manually controlled by individual pistol-grip, switches individual pistol-grip switches with LOWER, LOWER, OFF, and RAISE positions, with spring return return to OFF. One safety blade may be selected to be controlled by by the automatic automatic level control system. The The position of each safety blade is indicated indicated by separate digital read-outs, and the indicator by separate indicator lights on the console show when each each drive is at its "'IN" "IN" or "OUT" "OUT" limit and when the blade magnets are engaged armatures.. Position engaged with the armatures Position indication is accurate indication accurate to +/-0O.02
+/-0.02 inches.
The safety blades will scramscram from any position when stationary stationary or during during withdrawal withdrawal and insertion. In the event of a scram scram the safety safety blade drives drives automatically their "IN" automatically run in to their "IN" limits. I7.3.7 7.3.7 Regulating Regulating Blade Control The regulating blade identical position blade has identical indication and "IN" position indication "OUT" limit indication. It is "IN" and "OUT" manually controlled by manually by a separate separate pistol-grip pistol-grip switch and may be controlled controlled byby the automatic automatic level system. level control system. UWNR UWNR Safety Analysis Report Rev. 2 7-8 78Sp.20 7-8 2008 Sept. 2008
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- 7.3.8 Transient Rod Control Manual Manual movement of the transient rod drive is controlled controlled by the console-mounted, console-mounted, switch/light, push buttons which not only control movement, but also indicate in and out limits. Position Position indication is accurate to 0.02 inches. The transient rod drive may be selected indication selected to be controlled by the automatic level control system.
Air pressure is used to fire the transient rod to the selected position in pulse and square-wave square-wave modes and to engage and hold the transient rod at the drive position in manual and automatic modes. Additional lighted push-button push-button switches are installed to supportsupport these functions.
"READY/FIRE T ROD" indicator/switch Illumination of the "READY/FIRE indicator/switch indicates that the permissives for firing are met and the rod has not been fired. Illumination Illumination of the "ENG'D/AIR" "ENG'D/AIR" indicator/switch indicator/switch indicates movement movement of the transient rod from the full-in rest position as a result result of air having been applied. (Applying (Applying air while the transient rod drive is in the full in position causes the piston in the drive to move upward by compressing compressing the spring inside the shock shock absorber). Pressing Pressing the "ENG'D/AIR" "ENG'D/AIR" switch removes the air from the drive, causing the transient rod to drop to the full-in rest position.
Since the transient transient rod control is capable capable of introducing step changes changes in reactivity up to the Technical Specification limit, logic circuits are provided Technical Specification provided to assure the transient control rod is fired only under the appropriate appropriate conditions.
- 7.3.9 Automatic Automatic Level Control System The servo amplifier System amplifier (level controller) controls controls reactor power level in automatic and square-wave modes. The servo amplifier output drives either the regulating blade, transient rod, or a safety blade as selected by a servo element selector selector switch on the console. Only one element at a time may be selected; selected; selecting another replaces the previously previously selected element. The servo amplifier amplifier responds to a power level signal from one of the safety responds safety channel channel picoammeters picoammeters and controls speed speed and direction of the servo element through a servo motor.
In automatic automatic mode, the automatic level control channel uses period information from the Log N
- period channel to limit control element withdrawal withdrawal to maintain maintain a period period longer than a pre-selected selected level. In square-wave square-wave mode a servo error circuit is employed. This circuit allows servo operation only when the servo errorerror is less than 5%. Servo error (the difference between between scheduled power in % and the picoammeter indication scheduled indication in % of full scale) scale) is indicated at the console in both "square-wave" "square-wave" and "automatic" modes.
Additional Additional indicators indicators on the reactor control console are provided to indicate indicate "AUTO ON" and scheduled power.
- UWNR Safety Analysis Report Rev. 2 7-9 7-9 2008 Sept. 2008
7.4 Reactor Scram Circuits Protection System Reactor Protection The scram circuit, which initiates initiates shutdown shutdown by dropping the shim-safety shim-safety blades and the transient
- rod, is shown in Figure 7-2.
7-2. Scram is accomplished de-energizing the scram relays under accomplished by de-energizing under one of the following conditions:
- 1. Manual scram;
- 2. Fuel temperature temperature above above LSSS; 3.
- 3. Power level greater than 1.25 P max;
- 4. High voltage failure in control console; 5.
- 5. Loss of control power;
- 6. Coolant Coolant temperature temperature at core coolant entrance above 130°F;130'F;
- 7. Pool water level high or low.
In addition to these scram functions, the transient control rod logic includes a timer which causes
- the transient rod to drop to the full in position within 15 seconds seconds after the transient rod has been been fired in pulse mode only.
The key-operated key-operated console MASTER switch, designated 4S2 on the figures, is a cam-operated console MASTER cam-operated switch switch with a large number number of contacts which are selectively selectively operated in the three switch positions. (OFF, ON, and TEST). Six different different contact contact sets must be closed closed (and are closed only only in the "ON" "ON" position of the switch) in order order to reset the scram relays and apply power to the trip trip amplifier which supplies DC voltage to the shim-safety shim-safety blade magnets. A normally-open normally-open contact contact of Relay 6KIA 6K1A in the transient rod logic circuit opens when the relay is de-energized, de-energized, resulting in the drop of the transient transient rod. I Normally-open contacts of both 6K Normally-open 6KIA1A and 6KIB 6K 1B are in series with the alternating current power supply supply to the trip amplifier as shown in Figure 7-3. 7-3. Magnet power is turned off if if either or both of the scram relays de-energizes. de-energizes. UWNR u WNR Safety Safety Analysis Analysis Report Rev. 22 Report Rev. 7-10 7-10 Sept. 2008 Sept. 2008
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- UWNR Safety Analysis Report Rev. 2 c:::::
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7.5 Engineered SafetySafety Features Features Actuation Systems
- 7.5 Engineered Actuation Systems There are no engineered There engineered safety features actuation systems.
features actuation 7.6 7.6 Control Console Control Console and Display Instruments and Display Instruments 7.6.1 7.6.1 Alarm Alarm and Indicator System Indicator System When When an abnormal condition develops, abnormal condition develops, an audible audible signal sounds and a lighted annunciator annunciator begins begins to flash rapidly. The operator operator may press press the acknowledge acknowledge button to silence the audible signal, signal, at which time the audible audible signal stops and the lighted annunciator goes to steady lighted annunciator steady illumination. illumination. When condition is corrected, When the condition corrected, the lighted annunciator annunciator goes goes to slow slow flash, and the light light is extinguished extinguished when the operator operator presses presses the Reset button. The following conditions will actuate following conditions actuate the alarm alarm system: 1.
- 1. High area radiation level (also gives an alarmalarm at UW UW Police Department and Police Department initiates building evacuation);
building evacuation);
- 2. High experimental experimental facility radiation radiation level; level; 3.
- 3. Radiation Radiation monitor monitor failed low;
- 4.
5. 6. Evacuation Evacuation Alarm Alarm in Local; Stack Air particulate Local; particulate or gaseous activity CAM Air particulate particulate or gaseous activity above above normal level;
- 6. CAM gaseous activity activity above normnal normal level; 7.
- 7. Trouble in stack or continuous continuous air monitors; 8.
- 8. Neutron Neutron flux exceeding 1.15 times the normal value; exceeding 1.15 value; 9.
- 9. Reactor period less than a preset preset level; level; 10.
- 10. Count rate rate on startup counter counter approaching approaching saturation saturation level; level; 11.
- 11. Any scram;
- 12. Safety Safety blade blade disengaged disengaged from magnet; 13.
- 13. Failure Failure of high voltage voltage power power supply;
- 14..
14 Loss of Off-Site Power; UWNR Safety Safety Analysis Analysis Report Report Rev. 2 7-13 7-13 Sept. 2008 2008
- 15.
- 15. Fuel element temperature high; 16.
- 16. Core inlet temperature Core inlet temperature above above preset preset level; level; 17.
- 17. Cooling system temperatures temperatures above preset level; above preset level; 18.
- 18. Intermediate Intermediate coolant system low pressure.
coolant system 19.
- 19. Thermal Column Column door open;
- 20. stair actuated Chain switch across stair actuated or entry entry to High Radiation Radiation Area; 21.
- 21. Hold tank full;
- 22. Water level in pool two or more inches above or below normal (also gives an alarm at UW Police Department);
7.6.2 Indicator Lights To provide operating information for the reactor reactor operator, the following indicator indicator lights are provided: 1. 2. 3. 3. Scram reset; Safety blade magnet engaged; Power Power on;
*
- 4. Control Control elements elements in (distinct light for each);
5.
- 5. Control elements.out Control elements out (distinct light for each);
- 6. Automatic Automatic control control on; 7.6.3 Pneumatic Tube Tube System Panel The pneumatic pneumatic system system control control panel provides indication indication lights for system on, system purge in in
- progress, progress, isolation valves open, and rabbit in reactor. Buttons Buttons allow the console console operator operator to start and stop the system, and emergency emergency return a rabbit sample. The pneumatic pneumatic system is described described in section 10.2.3.
section 10.2.3. UWNR U W N R Safety Satety Analysis Report Rev. 2 7-14
/-14 Sept. 2UooS 2008
- 7.6.4 Ventilation System Panel
- The ventilation system control panel provides indication for both of the main exhaust fans (EF-7 and EF-8), the air handling unit (AHU-5), the fume hood exhaust fan (EF-13),
& Thermal Column & Column exhaust fan (EF-17). Each fan has separate is running and whether (EF- 13), and the Beam Port separate lights indicating whether whether the fan whether the fan has power available to run if needed. All fans except the fume hood exhaust fan (EF-13)
(EF-13) include switches for operation. Digital indications are provided for stack exhaust exhaust flow-rate (scfm), (scfm), differential differential pressure pressure (inches of water water column) across the mainmain exhaust filter bank, and static duct pressure pressure (inches (inches of water column) in the Beam Beam Port & & Thermal Column exhaust Column exhaust duct before the exhaust fan EF- 17. EF-17. The ventilation ventilation system is described described in section section 9.1. 9.1. 7.6.5 System Panel Cooling System Operating controls and indicators at the control console for the cooling system include switches Operating to operate intermediate, chilled water, and diffuser pumps (the hydraulic irradiation operate primary, intermediate, irradiation facility pump starts when the diffuser pump starts) along with indicatorsindicators which light when the pumps have discharge discharge pressure. In addition, a pressure switchswitch on the intermediate coolant intermediate coolant system provides provides indication that the system is pressurized. pressurized. The cooling system is described in chapter 5. chapter 5. 7.6.6 Whale System PanelPanel
- The hydraulic hydraulic irradiation facility (whale system) includes indicator lights for flow direction, sample in, and buttons to reverse flow direction, as well as indication discharge pressure.
7.7 pressure. The whale system is described in section Radiation Monitoring Monitoring Systems indication that the whale pump has section 10.2.4. 7.7.1 Area Radiation Radiation Monitors The radiation monitors monitors are arranged arranged into three systems; the primary area monitors, experimental experimental facility area monitors, and air activity monitors. The primary primary area monitors are located as follows: 1. I. Demineralizer area; Demineralizer
- 2. On the reactor bridge about one foot above the water surface; 3.
- 3. Beside the thermal column door;
- 4. In the control console console area.
All Area Radiation monitor units have ranges from 0.1 to 10000 mr/hr. mr/hr .
- UWNR Safety Analysis UWNR Analysis Report Report Rev. 2 7-15 7-15 Sept. 2008 2008
Unit 1 supplies information infonnation on radiation level from the demineralizer. It is set to alarm alann at a radiation level just above that expected nonnal run. Unit 2, located just above the pool water expected in a normal water
- alarms at ~a radiation level just above that reached during full power operation. Unit 3 is level, alanns located beside the thermal thennal column. It too is set to alarm alann just above normal nonnal operating operating level. This unit will give an alann alarm if the thermal thennal column door is left open when the reactor is operatedoperated at any substantial substantial power. Unit 4 indicates the dose rate in the console area. The 4 units indicated above are connected to the Reactor Laboratory evacuation alarm. An alann evacuation alann. alarm from one of these units will sound the evacuation evacuation alarm alann if it is not corrected ifit corrected by the operator within 30 seconds (See (See procedure procedure UWNR 150). 150).
The Experimental Experimental Facility Facility Area Radiation Monitor Monitor is an area radiation monitor system installed to preclude the possibility of unknowingly unknowingly generating high radiation radiation levels by operating operating the reactor at high power levels with the beam ports open, or by return of an intensely intensely radioactive pneumatic tube sample. The sensors for this system are installed on the walls of the Reactor Laboratory in direct line with the beam ports and at the pneumatic tube send-receivesend-receive station. The system gives visual and audible alannsalarms at the console if the radiation radiation level exceeds a preset value. The pneumatic tube monitor also provides local alann alarm and indication. The monitors are nonnally normally set to alarm at a radiation level equivalent equivalent to a dose rate of 50-100 mrem/hr at the beam port flange (10 (10 mremlhr mrem/hr at the detector location). location). The pneumatic pneumatic tube monitor also is normally set set to 10 mrem/hr, but the pneumatic tube operating procedure states that it may be set to a higher higher level if calculated sample activity is expected expected to result in a higher reading. 5 7.7.2 Stack Air Monitor Monitor The stack air monitor measures both particulate sealed gas proportional particulate and gaseous gaseous activity of the air discharged from the stack. Particulate activity is collected on filter paper and countedcounted with a thin end-window end-window proportional tube. Gaseous activity is also measured with a gas proportional tube.
- operates by detecting p The system operates P activity. Both particulate particulate and gaseous activity levels are
- recorded, recorded, and provide provide annunciation annunciation should preset levels be exceeded. In addition, gaseous activity levels are integrated to provide a record of total gaseous activity discharged from the stack.
The sensitivity sensitivity of the particulate particulate activity monitor allows detection of concentrations concentrations of about 1.OE-10 !lC/ml 1.0E-I0 material with a single p [iC/ml of a material P3particle emitted emitted per disintegration. efficiency is disintegration. The efficiency higher if more than one p P3particle is emitted per disintegration. The sensitivity sensitivity of the gaseous activity monitor monitor is such that a concentration concentration of about 1.0E-6 1.OE-6 !lCi/ml [LCi/ml of Ar 41 at the stack discharge Ar41 efficiency varies with the number of p can be detected by the instrument. The efficiency P3particles emitted by the isotope being detected. The primary activity expected expected to be present in the stack discharge is Ar41 activity. An identical instrument, provided as a backup for the Stack Stack Air Monitor, is operated as a Continuous Continuous Air Monitor. It samples the atmosphere immediatelyimmediately above the surface of the reactor pool, although it can be made to sample other locations locations when desired. The backup air activity
- monitor can be connected to the stack monitor flow path, should the stack monitor fail.
UWNR UWNR Safety Analysis Report Rev. 2 7-16 7-16 2008 Sept. 2008
7.8 References
- 7.8 References There are no references for this chapter.
There
- UWNR Safety Analysis Report Report Rev. 2 7-17 7-17 Sept. 2008 2008
- This page is intentionally left blank.
- UWNR UWNR Safety Safety Analysis Report Report Rev. 2 7-18 7-18 Sept. 2008 2008
*
- 88 ELECTRICAL POWER ELECTRICAL POWER SYSTEMS SYSTEMS 8.1 8.1 Normal Electrical Normal Electrical Power Power Systems Systems There electrical power supplies There are no electrical supplies that are critical for maintaining maintaining the facility in safe shutdown, even for extended extended periods periods of time.
The Reactor Reactor Laboratory's Laboratory's electrical electrical power originates in room B 11 power originates I 10011 of the Mechanical Mechanical Engineering Engineering Building. The entire system is shown shown in Figure 8-1. 8-1. Utility power to the building is supplied by circuits 1421 and 1422. by circuits 1422. Below is a description description of each electrical circuit used in the electrical circuit Laboratory. Reactor Laboratory. 8.1.1 8.1.1 277/480 VAC 277/480 VAC 33 Phase Electrical Power Power 277/480 VAC 277/480 VAC is provided provided to the Reactor Laboratory by Reactor Laboratory by a 3000 3000 kVA kVA transformer transformer and through a ground fault interrupted 3000 A interrupted 3000 A circuit circuit breaker. The Laboratory Laboratory itself has its own 225 225 AA ground ground interrupter and circuit fault interrupter circuit breaker (08). This circuit breaker (#8). circuit connects to the Reactor Reactor Laboratory's Laboratory's 277/480 panel 277/480 (4R1) which panel (4Rl) located behind which is located Console just outside behind the Reactor Console outside of the Reactor Reactor Control Room. panel supplies 480 VAC This panel N- 16 Diffuser and Laboratory's N-16 VAC to the Reactor Laboratory's Whale Pumps. and Whale
- Although Although both of these pumps are on the same same circuit, each has its own motor starter. starter. The Diffuser located immediately Diffuser Pump starter is located immediately behind behind the Reactor Reactor Console Console in the Reactor Control Room Room and the Whale starter is located to the right of the Diffuser Whale Pump starter Diffuser Pump starter.
starter. The The Diffuser Diffuser Pump also has a local disconnect disconnect which is located located next to the Diffuser Pump Pump on the Reactor's Reactor's Shield Step. The Diffuser Diffuser and Whale Pumps can be turned turned on and off off using a single shared toggle toggle switch located graphic mimic panel located on a graphic panel in in the Reactor Control Console.Console. 277/480 panel also provides The 277/480 power to the 33 loops of the provides power the Reactor Reactor Laboratory's Laboratory's Cooling Intermediate and Chilled) System. Each loop (Primary, Intermediate disconnect switch Chilled) has its own local disconnect switch located just below the Reactor's Reactor's Control Control Room Room and each can be turned on and off using individual toggle switches located individual graphic mimic located on a graphic mimic panel panel in the Reactor Control Console. Reactor Control Console. 277/480 panel also supplies power to the Reactor Laboratory's Finally, the 277/480 Laboratory's overhead crane and to overhead crane the primary windings windings of the Reactor Laboratory's two power Reactor Laboratory's power transformers, transformers, Tl TI and T2. TheseThese transformers are described in detail below. 8.1.2 8.1.2 277/480 VAC 33 Phase 277/480 VAC Phase Backed Backed Up Up Electrical Electrical Power Power The The Reactor Laboratory's 277/480 Reactor Laboratory's 277/480 VACVAC backed up electrical electrical power power is supplied supplied by by the the same 277/480 277/480 VAC VAC 3000 3000 kVA kVA transformer transformer and through the same 3000 3000 A circuit breaker A circuit breaker as the Laboratory's 277/480 Laboratory's 277/480 VAC 3 VAC 3 phase normal electrical normal electrical power. This bus has has its 1600 A own 1600 A breaker and passes through an Automatic Automatic Transfer (ATS). This ATS Transfer Switch (ATS). ATS automatically automatically
- Safety Analysis UWNR Safety Analysis Report Report Rev. 2 8-1 81Sp.20 Sept. 2008 2008
switches the bus's source from utility to the Mechanical Mechanical Engineering Engineering emergency emergency diesel generator.
- After the ATS, A TS, this circuit passes passes though another another 1600 A breaker breaker and a 100 A breaker to a circuit circuit within the Mechanical Mechanical Engineering Engineering Building Panel Panel4CRl. located in room 1133 4CRl. This panel is located 1133 Mechanical Engineering. A circuit in 4CR1 powers the Reactor Reactor Laboratory's Laboratory's backup/makeup backup/makeup air air compressor. When the Reactor Laboratory's compressed compressed air system, provided by the Mechanical Mechanical Engineering Building, Engineering Building, falls below 75 psi, this local air compressor compressor activates.
The Reactor Reactor Laboratory's ventilation system is also powered by this bus. Bus EMCC6 is Laboratory's ventilation connected connected to the Reactor Laboratory's 277/480 VAC VAC backed up electrical power through two 400 A fuses. Bus EMCC6 powers powers the Reactor Reactor Laboratory's two ventilation fans, EF-7 and EF-8,EF-8, as well as the Reactor Reactor Laboratory's Laboratory's air handling handling unit, AHU-5. AHU-5. Finally, the Reactor Laboratory's Support Support Space basement emergency emergency lights are also powered powered by by the 277/480 VAC backed up electrical electrical bus through panel paneI4CR2. 4CR2. 8.1.3 240 24044 wire VAC Electrical Electrical Power, Transformer TI T1 24044 wire VAC is provided 240 provided by transformer transformer TI. T1. This transformer's primary winding is is connected to 480 VAC from panel connected panel4RI 4RI and its secondary secondary winding feeds panel R. Both panel R T I are and T1 are located behind the Reactor Console just outside of the Reactor Control Room. Panel R supplies power to the Thermal Column Door motor which allows the Thermal Panel Thermal Column Column to be opened opened and closed closed by pressing pressing a pair of buttons located on the Thermal Column Column Door itself. itself.
- In addition, Panel R supplies 240 V VACAC power power to several NEMA "twist-lock" type outlets located on the Reactor Reactor Shield at each beam port.
8.1.4 208/120 3 wire V VAC Electrical Power, Transformer AC Electrical Transformer T2 2081120 208/120 VAC is provided by transformer T2. This transformer's transformer's primary winding is connected connected to 480 VAC from Panel4RI Panel 4RI and its secondary secondary winding feeds Panel Panel 2. Both Panel 2 and T2 are located behind the Reactor Console just outside of the Reactor Control Room. Panel 22 supplies Panel supplies power power to to aa pressure pressure makeup makeup pump pump inin the the Intermediate Intermediate Loop of the Loop of the Reactor's Reactor's Cooling System. It also supplies power to the Demineralizer Demineralizer Pump in the Reactor's Primary through an outlet located on the west wall by the Demineralizer. Coolant Makeup System through Demineralizer. In In addition, Panel 2 powers the circulating pump of the Waste Holdup Holdup System. Various 120 VAC outlets throughout throughout the lab as well as a set of 120 VAC and 208 VAC outlets at each beam port are also powered powered by Panel 2. Finally, though the use of an uninterruptible uninterruptible power supply (UPS), the Reactor's Reactor's Control Control Console is powered through Panel 2. See section 8.2 for details of this UPS. All power to the Control Console is 120 V VAC. AC. UWNR Safety Analysis Analysis Report Report Rev. 2 8-2 8-2 2008 Sept. 2008
*
- 8.1.5 277 V AC Lighting Electrical Power VAC Power Most Reactor Laboratory Laboratory lighting is supplied supplied by the same 277/480 VAC 3000 3000 kVA transfonner transformer and through the same 3000 A circuit breaker Laboratory's 277/480 33 phase electrical breaker as the Laboratory's power. The lighting circuit has its own 200 A breaker breaker and ground fault interrupter interrupter circuit. This circuit is connected through a cutoff switch (located in room B B 1140 1140 Mechanical Mechanical Engineering)
Engineering) to Panel BlB5 BI 135 (located in room BI BlB5 Mechanical Engineering). Panel BlB5 135 Mechanical BI 135 has various breakered circuits which then power the various lights within the Reactor Laboratory breakered Laboratory and the Reactor Laboratory's support space. Reactor 8.1.6 277 V AC Backed VAC Backed Up Lighting Electrical Power A few selected selected lights within the Reactor Reactor Laboratory Laboratory and the support space are powered by a separate 277 VAC circuit. These lights are always on and are connected through an Automatic separate Transfer Switch (ATS) which transfers the circuit from utility power to an emergency generator generator in the event of a power power failure. 8.2 8.2 Emergency Electrical Power Systems Emergency Electrical Neither of the systems described in this section are required or necessary necessary for safe facility operation or shutdown shutdown..
- There There are several Laboratory Emergency Electrical several Emergency Laboratory utilizes. Those which by a diesel 800 kW 277/480 The other Emergency Emergency Electrical Electrical Power Systems, shownshown in Figure 8-2, which the Reactor which are a part of the Mechanical Mechanical Engineering 277/480 generator located in room Bl002A Reactor Engineering building are powered B 1002A Mechanical Mechanical Engineering.
Uninterruptible Power Electrical Power Systems are two battery based Uninterruptible powered Power Supplies located located within the Laboratory Laboratory in rooms 12151215 and B 1135E Mechanical BlB5E Mechanical Engineering. 8.2.1 . Reactor Laboratory Laboratory Uninterruptible Power Supply A 6000 VA Uninterruptible Power Supply (UPS) is connected between utility power V A Uninterruptible power and the reactor's control console. This system is installed solely to protect the reactors reactors instrumentation instrumentation from sudden power losses as well as dirty power. This UPS provides provides protection protection from surges, brownouts, noise, spikes, frequency variations, transients and harmonichannonic distortion. In the eventevent of a loss of utility, the UPS is capable of providing up to 76 minutes of power to the reactor's reactor's control console control console and instrumentation. This allows allows for a monitored shutdown shutdown and facilitates facilitates uninterrupted recording and monitoring. A failsafe SCRAM and alann alarm connected connected to a non non backed up power circuit circuit ensures that control blades drop and the operator operator infonned the is informed in event event of a loss of utility. Even Even with a UPS failure the reactor is designed to SCRAM in the event of a loss of power. Therefore, this UPS is not required required for a failsafe shutdown when power is lost and in the event of of
- UWNR Safety Analysis Report Rev. 2 8-3 8-3 Sept. 2008 2008
UPS failure a UPS I8.2.2 8.2.2 failure and loss Imaintaining loss of maintaining the facility of utility, the facility in Reactor Laboratory the reactor will shutdown. This UPS in a safe shutdown shutdown condition, even for Laboratory Support Space Uninterruptible UPS is not required for for extended extended periods Uninterruptible Power Supply periods of time.
- IA 6000 V A 6000 VA Uninterruptible Power Supply (UPS)
A Uninterruptible (UPS) identical identical to the Reactor Reactor Laboratory's Laboratory's is located in the Laboratory's Laboratory's support space. space. This UPS UPS serves as both an installed spare for the Reactor ILaboratory UPS and as a support space power backup. Laboratory UPS backup. In the support space, UPS powers space, this UPS several outlets, allowing the support space's counting computers, sample changing changing equipment
- Iand, and, most importantly, the Stack the Stack Air Monitor (SAM)
(SAM) and Continuous Air Monitor (CAM) to Monitor (CAM) Icontinue continue operating operating or be shut down gracefully in the event of a utility power failure. I8.2.3 8.2.3 Mechanical Mechanical Engineering Engineering Building Building Emergency Emergency Power Power IAs As mentioned mentioned in section 8.1 8.1 above, the Mechanical Engineering building Mechanical Engineering building has an emergency emergency emergency power systems. These systems include the backed up backup generator and several emergency Femergency emergency lights panel, which powers selected lights throughout the Reactor Laboratory Laboratory and the IMechanical Mechanical Engineering Engineering Building, the backed up equipment panel, panel, which powers powers the Reactor Reactor FLaboratory's Laboratory's air compressor, compressor, smoke detector/fire detector/fire alarm and ventilation. ventilation. As mentioned in Section I8. 1, these systems are connected 8.1, connected through automatic transfer switches through automatic switches which switch them to Fgenerator generator power in the event of a loss of utility power. IIn In addition, the Mechanical Engineering Buildings Mechanical Engineering Buildings fire suppression suppression system pump also also has its own
- Iautomatic automatic transfer transfer switch and emergency emergency power. This system system includes includes the fire suppression suppression Isystem I
system in the Reactor Laboratory UWNR UWNR Safety Safety Analysis Analysis Report Report Rev. Rev. 2 84Sp.20 8-4 Sept. 2008 2008
- c ....~
* *
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- Figure 8-2 Backup Electrical System UWNR Safety Analysis Report Rev. 2 8-6 Sept. 2008 Sept. 2008
- 99 AuxiLIARY SYSTEMS AUXILIARY SYSTEMS 9.1 9.1 Heating, Ventilation, Ventilation, and Air Conditioning Systems The heating, ventilation, and air condition system (hereafter called the ventilation system), is a dedicated reactor laboratory interaction with the rest of the building. The system laboratory system with no interaction incorporates a fresh air supply air handling unit, two exhaust fans, filter banks, variable supply incorporates and exhaust air mixing boxes, and duct work. The system provides HVAC HV AC service to both the reactor laboratory laboratory and auxiliary support support spaces. Figure 9-19-1 shows the ventilation ventilation system flow flow schematic, schematic, while Figure 9-2 9-2 and Figure 9-3 show the physical location of ventilation systemsystem components.
The ventilation system is designed to prevent the spread spread of airborne particulate radioactive airborne particulate radioactive particulates with high-material into occupied areas outside the Reactor Laboratory. It removes particulates particulate activity are efficiency filtration and assures that all releases of either gaseous and particulate monitored elevated release point. Accidents which might result in discharge monitored and discharged at an elevated discussed elsewhere in this report, and remarks may be material from the stack are discussed of radioactive material concentrations which might be expected. In addition, a portion of the found there indicating the concentrations ventilation system vents the beam beam ports, thermal column, and liquid waste holdup tank to assure that air flow is from the Reactor Laboratory Laboratory into these facilities. 9.1.1 Air Handling Handling Unit
- floor of the Mechanical Mechanical Engineering Engineering building access. The unit supplies 9150 building behind (AHU-5) mounted in the fifth The ventilation system consists of a dedicated air handling unit (AHU-S) behind a security fence which only reactor staff can 70'F fresh air at a relative humidity of 30%
91S0 scfm 70°F 30% to the reactor reactor space through 7 variable air volume boxes; S5 in the auxiliary laboratory and auxiliary support space support space, 1 in the reactor reactor laboratory confinement, reactor control room. The confinement, and 1 in the reactor variable frequency drive which maintains duct static pressure unit is powered from a variable pressure at 1.0 inches interlocked with the exhaust fans to trip if neither exhaust fan is of water gauge. The unit is interlocked running. runmng. 9.1.2 Exhaust Fans mounted exhaust fans (EF-7 ventilation system incorporates two roof mounted The ventilation (EF -7 and EF EF-8),
-8), each each individually capable of exhausting 9600 scfm air. Normally only one fan is running, but in emergency venting mode both fans may be run for increased emergency increased exhaust dilution with 19200 scfm scfm reactor laboratory air. The reactor -1.5 inches of water gauge to laboratory exhaust duct pressure is maintained at -1.S maintain air flow from surrounding areas into the reactor laboratory.
laboratory. If one exhaust fan fails, the other fan attempts to pickup. If both exhaust fans fail, an interlock other interlock will trip the air handling unit to prevent laboratory.. prevent having a positive pressure in the reactor laboratory
- UWNR Safety Analysis Report Rev. 2 9-1 9-1 Sept. 2008
Each fan has isolation dampers that close when when the fan is not running running to prevent short cycling cycling the
- Iexhaust exhaust flow path. Both fans take take a suction on a common header (located (located below the roof on the fifth floor Mechanical Mechanical Engineering)
Engineering) whichto which all exhaust variable variable air volume boxes discharge discharge to Ithrough through the filter bank. This provides diluting air from the auxiliary support space auxiliary support space prior prior to Idischarging reactor confinement discharging reactor confinement air. In addition, the common header common header equippedis equipped with duct work Iwhich which draws a suction suction on outside outside air through the reactor reactor attic for further dilution. This outside air is not filtered. This duct work includes an automatically automatically controlled controlled damper damper which which is adjusted adjusted to maintain maintain the exhaust exhaust duct pressure -1.5 inches of water gauge. This is necessary pressure at -1.5 necessary because because Ithe the exhaust exhaust fans do not have variable frequency drives and so must always run at full variable frequency full capacity. An additional manually controlled available for bringing controlled damper is available bringing in additional additional outside outside air into Ithe the common header, but this damper is normally normally locked closed. closed. The air from this common header is continuously header monitored by continuously monitored by the the stack stack air monitor to maintain maintain a record of radioactivity radioactivity discharged. indication is also displayed in the control discharged. Flow rate indication control room. IThe The exhaust fans are at a height of 26.5 26.5 meters above grade. These fans, manufactured manufactured by by IStrobic, Strobic, incorporate incorporate a unique nozzle design which operates on which operates on a principle principle of internal and externalexternal Iexhaust exhaust stream stream dilution. dilution. The nozzle entrains outside air with the primary exhaust exhaust stream stream to Iproduce produce a substantially substantially diluted exhaust exhaust stream. This enhancedenhanced flow streamstream then then undergoes undergoes a . Ipressure pressure increase increase to increase increase stack outlet velocities velocities which which increases increases the effective stack height. effective stack IHowever, However, the analysis Appendix A analysis in Appendix A neglects the increase increase in stack stack height height for a bounding bounding Ianalysis. analysis. I9.1.3 9.1.3 IAll Idilution). Filters All ventilation ventilation exhaust exhaust is filtered just prior dilution). This main filter bank (F Ifollowed by 16 nuclear followed by 16 nuclear grade
-21)
(F-2 HEPA HEPA prior to entering filters entering the common
- 1) consists of a 4x4 array which which are rated common header (before any outside array consisting at consisting of 16 99.97% efficiency 99.97%
16 pleated efficiency for outside air pleated pre-filters pre":filters particles of
- particles of 1 I0.
0.11 micron micron and larger. The pneumatic pneumatic system system fume hood also has its own basement basement filter bank
- 1 even though the fume hood exhaust is sent to the main ventilation (F-22), even ventilation exhaust system. This 1 Ifilter filter bank has 1 pre-filter pre-filter followed followed by by 1I HEPA HEPA filter located upstream of the fume hood exhaust located upstream exhaust 1 Ibooster (EF- 13). The main~cated booster fan (EF-13). main filter bank (F-2 1) is located with the common header header in the fifth 1 Ifloor Mechanical E floor Mechanical Engineering, n g i n e e r i n g , _ .. The fume hood exhaust exhaust basement basement filter 1 bank (F-22)
(F-22) is located in the auxiliary supportsupport space in room Bl135A. Bi 135A. The pressure pressure drop for the 1 main filter bank bank is indicated indicated in the control room. Local indications indications for pressure across the pressure drop across 1 Ipre-filters pre-filters and across the HEPA HEPA filters are on the fifth floor Mechanical Mechanical Engineering Engineering and and the 1 Ibasement. basement. 1 1 I9.1.4 9.1.4 Exhaust Variable Volume Boxes Variable Air Volume I 1 IAsAs shown in Figure 9-1, 9- 1, the exhaust system incorporates incorporates 10 10 variable variable air volume boxes. Four located in the auxiliary boxes are located auxiliary support support space, space, three three are located located in the reactor reactor laboratory laboratory for Inormal 1 I normal venting, and three are located located in the reactor laboratory laboratory for emergency emergency venting. Of Of the I Ifour four boxes in the auxiliary support space, space, one is designated designated for the pneumatic pneumatic system fume hood. UWNR Safety Analysis Report Rev. 2 UWNR 9-2 9-2 Sept. 2008 2008
*
- Each box is adjustable adjustable from 0%
0% to 100% 100% design flow. flow. The normal confinement confinement boxes have a combined rating of 2700scfm. With this flow rate, assuming a confinement confinement volume of 200Gm 3,, it of2000m would take 26.16 minutes (1 569s) to completely (1569s) completely exhaust confinement. 9.1.5 Emergency Venting Mode Emergency Emergency Emergency venting mode is for use when it is desirable to rapidly change the air in the reactor reactor laboratory laboratory to prevent contamination to adjacent occupied areas. Use of emergency prevent spread of contamination emergency venting mode is governed governed by the Emergency Emergency Plan, which states that the decision to operate in emergency emergency venting mode should be reached by common common consent consent of the Emergency Coordinator Emergency Coordinator and the campus campus Health Physics organization. Emergency Emergency venting venting is initiated by one of two large red push-buttons (with switch covers) located in the control room by the south door and in the first floor vestibule, room 1200J, 12001, by the east catwalk catwalk door. When either of these buttons are pressed, the second exhaust fan activates to double the flow-rate. The three variable air volume boxes designated for emergency emergency venting mode are normally closed until emergency emergency venting mode is activated by either of the switches. Because the negative negative pressure pressure in reactor confinement confinement would be extreme extreme with both fans running, a makeup exhaust damper in the confinement confinement roof roof opens in emergency emergency venting venting mode to allow outside air to enter confinement confinement through the reactor attic. This makeup exhaust damper is normally closed with a weather-proof weather-proof seal. 9.1.6 Beam Port and Thermal Column Ventilation System System
- The Beam Port and Thermal Column Ventilation system is designed to sweep out the Ar-41 activity activity present in an experimental facility when the facility is opened. During ordinary operation operation the experimental experimental facilities are closed and there is an essentially discharged. The average concentration essentially zero rate of discharge.
When a beam port flange or the thermal column door is opened there is a slug of activity discharged will, therefore, concentration discharged therefore, be extremely low due to dilution by the rest of the ventilation ventilation system and the fact that no activity discharged most ofthe activity is discharged of the time. Section 11.1.1.1 of this report discusses the levels of activity discharged. The Beam Port & & Thermal Column ventilation system consists of a booster exhaust fan (EF-17) (EF -17) mounted in the Reactor Attic (room 3110), 3110), which discharges into the main ventilation ventilation system system exhaust duct work (before the main filter bank F-21), F-2 1), and a makeup damper located in the Reactor Reactor Laboratory. The booster exhaust fan is needed to provide sufficient suction on the Beam Port && Thermal Column ventilation system. The makeup damper is adjusted adjusted to maintain approximately 960 cfm at 1.5 inches suction pressure approximately pressure and an air velocity of about 40 feet per minute into all beam ports and the thermal column, should all be opened simultaneously. Normal Normal flow rate with the system sealed is about 450 cfm. The thermal column shielding shielding door is weather-stripped weather-stripped to maintain a nearly nearly airtight seal. An air-operated air-operated flapper valve at the end of the duct connected to the thermal column column and beam port vents is normally open. This maintains a slight negative pressure pressure within the thermal thermal column column when the door is closed, but prevents the full static suction of the system from forcing air in through the thermal thermal column door seals. When the thermal column door is opened, however, the flapper valve closes and full system suction is
- UWNR Safety Safety Analysis Report Rev. 2 9-3 9-3 2008 Sept. 2008
impressed impressed on the thermal column to cause the desired in-flow of air. In addition, a ball check
- valve within the thermal column vent prevents mixing of the air in the thermal thermal column column with the ventilation flow except when ventilation when the door is not sealed.
sealed. The facility ALARA ALARA program identified this program identified feature as one which resulted significant reduction resulted in a significant reduction in activity activity discharged through the stack stack increasing the Ar-4 without increasing Ar-411 dose within the laboratory. laboratory. Also incorporated incorporated into the system for ALARA considerations ALARA considerations are ball-check ball-check valves in the vent line from each beam port. Because the beam'port. pressure drop in the BP&TC BP&TC Ventilation System duct causes causes lower pressures pressures in beam ports closer closer to the exhaust exhaust fan, the common drain connection connection for the four beam ports allows allows air flow through through the drain connections connections from higher to lower lower pressure pressure beam beam ports. The check check valve prevents prevents flow into the beam port vents, vents, thus preventing circulation through the drain system and substantially preventing circulation substantially reducing reducing the amount amount of Ar-41 discharged discharged to the atmosphere. atmosphere. Should a beam port rupture and fill fill with with water, water water would leak out of the flapper valve at the end of the Beam Port & & Thermal Thermal Column ventilation ventilation system system on the shield step, which would would still leave at least 1111I ft water ft of water covering the core. The water would eventually eventually spill onto the reactor laboratory floor laboratory floor and and into into the the holdup holdup tank; tank; no no water water would leak out would leak out ofof the the ventilation system ventilation system outside of the reactor reactor laboratory because because of the height of the filter bank and exhaust exhaust fans. The vent of the waste holdup holdup tank is also connected connected to the Beam Port and Thermal Thermal Column Column system in order to assure Ventilation system assure any gaseous effluent from the holdup tank tank is discharged discharged through through a monitored monitored release path. In addition, the stack stack air activity discharges the activity monitor discharges 9.1.7 9.1.7 stream extracted sample stream Pneumatic extracted from the stack into this ventilation system. Pneumatic System System Fume Fume Hood The pneumatic system fume hood includes Exhaust Hood Exhaust includes its own exhaust booster fan (EF- 13) and filter bank (EF-13) bank
*
(F-22) to (F-22) to help help ensure ensure fume fume hood hood air air velocity velocity isissufficient sufficient toto protect protect the the operator operator in in the the event event ofof a a sample breakage. This booster fan exhausts into the common header of the main ventilation sample breakage. This booster fan exhausts into the common header of the main ventilation system... The system The fume fume hood exhaust fan hood exhaust fan isis automatically automatically activated activated upon upon starting the pneumatic starting the pneumatic system, or it may be manually manually run independent independent of the pneumatic pneumatic system if desired. Operation Operation of of the booster fan has a negligible impact on the main ventilation ventilation system system exhaust flow-rates flow-rates due to the small size of the booster booster fan relative to the main exhaust fans. The control room ventilation room ventilation system panel includes includes indication that the booster fan is running running and and that it has power power available available to needed, as well as the pressure drop across run if needed, across the fume hoodhood filter bank bank (F-22). UWNR Safety U"R Safety Analysis Report Report Rev. 2 9-4 9-4 2008 Sept. 2008
- e
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- Figure 9-1 UWNR Safety Analysis Report Rev. 2
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"-Ie z> El"lergency LoU Ve>ntlng Mode
~ake>up Dal"1pe>r Re>actor Lab ConFineMent BI135 B!l35A B1l35Cj ET Shop i
Sl.Ipply Supply BI135 Supply MlJ.chine Shop Si"jpply BllJ5D BI135 B!135B Supply HPGe , Supply o a R~ NC EMergency Ne El"lergency NC EMerge>ncy Roon Exhnust ROOM [)(ho.ust Roon Exhaust Roan Ex:haust o Supply E)(hnust [x:ho.ust Exhaust Rea.ctor lab Reactor Lab Beo.l"I Port Floor BI135
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*
- UWNR Safety Analysis Report Rev. 2 UWNR Safety Analysis Report Rev. 2 9-6 9-6 Sept. 2008 Sept. 2008 *
- Figure 9-3 Ventilation System Cross Section Ventilation Section Looking East
- UWNR Safety Analysis Report Rev. 2 9-7 9-7 Sept. 2008
*
- Figure 9-4 Fuel Storage Positions 0
9.2 9.2 Handling Handling and Storage of Reactor Fuel 9.2.1 Fuel Handling UWNR UWNR Safety Safety Analysis Analysis Report Report Rev. 2 9-8 9-8 Sept. 2008 2008
*
- 9.2.2 9.2.2 Fuel Storage New (unirradiated) reactor fuel can be handled manually and is stored in a steel safe. Those needing further details on new fuel storage should consult the facility security plan.
Sufficient Sufficient storage storage room is provided provided for all on-site irradiated fuel. The fuel storage locations are indicated in Figure Figure 9-4. All fuel storage storage facilities are designed to allow sufficient sufficient convective convective water flow to remove decay heat. 9.2.3 Fuel Bundle Maintenance Maintenance and Measurements Measurements
- Fuel bundle disassembly, assembly, and fuel element bow and elongation measurements conducted underwater conducted underwater installed)
. designed de .
using a special tool (shown in Figure Figure 9-5 with for both these purposes.s. This tool is used under a a du~ measurements are dummy element aboutb o u t _ The fuel-element handling tool also operates
. the socket and crowsfoot wrenches. For disassembly or assembly of fuel bundles, the operates maintenance maintenance and measuring measuring tool holds the bundle, provides a reference plate for a crows-foot crows-foot wrench, provides storage space for four individual individual fuel elements, and restrains the individual elements after they have been screwed screwed out of the bottom end box. An air-operated air-operated clamping device reproducibly positions the bottom end box of the fuel bundle. The crows-foot crows-foot wrench must be used for the initial loosening of each element (and for tightening tightening each element to the specified torque upon reassembly). Once the elements elements are loose enough to tumturn freely the top end fitting is removed so a socket wrench can be used on a hexagonal portion of the top fitting to completely unscrew the individual element from the bottom end box. The fuel element handling-completely tool then may be used to remove individual elements and place them in the storage positions.
While it is possible to disassemble disassemble and re-assemble re-assemble the fuel bundles, it is tedious even with use of the written written procedure and practice. Disassembly or assembly of an element takes about 30 minutes. Required measurements measurements of fuel element bow and elongation also are made with the maintenance and measurement measurement tool. Because of the excessive time and added handling required to ofthe disassemble disassemble the bundle and measure measure each element element in a separate measuring tool, the tool was
- UWNR UWNR Safety Analysis Report Rev. 2 9-9 9-9 2008 Sept. 2008
designed to make the measurements measurements without disassembly. Figure Figure 9-6 shows the three sensors
- employed (a portion of the housing is removed in this view.)
view.) Each uses a differential differential transformer transformer as a transducer to give a remote electrical output proportional proportional to displacement displacement of the sensors. The X and Y sensors employ spring-loaded aluminum aluminum wheels attached attached to the differential transformer cores. When the bundle is lowered into the tool the wheels are forced back and they transformer then ride on the fuel element clad surface. These sensors are adjusted to give a zero signal for a standard fuel element element dummy. The length sensor differential transformer is actuated by one lobe of a cam. A second lobe of this cam is rotated into contact with the top edge of the fuel element cladding cladding by a leaf spring attached to the operating rod. The cam pivots out into the measuring position only when the operating operating rod is fully withdrawn. The length sensor is also adjusted to zero output for the standard fuel element element dummy. A readout device is positioned on the pool curb or reactor bridge and connected to the underwater portion of the tool. Differential transformer transformer core position is indicated by a meter for length measurements, and a recorder output is provided provided to the horizontal axis of an X-YX -Y recorder for the bow measurements. measurements. Polarity is set so a~ an increase in element element length or a bow away from the center-line center-line of the bundle gives a positive meter indication indication or recorder readout. The dummy fuel bundle has one dummy elementelement exactly exactly 0.100 inches longer than the other three
- elements. This element element also has a section in which the radius has been reduced reduced by 0.060 inch, and and a section in which the radius has been increased increased by 0.060 inch. By using this element element the attenuation attenuation and zero controls on the readout box may be adjusted to give a calibrated calibrated readout of of bow and length.
After calibration calibration of the tool, measurement measurement can be 'made made on standard standard fuel elements. The standard setup used gives a 1 cm horizontal displacement for 0.060 inch transverse bend bend (bow) and a meter reading of length in thousandths of an inch 'deviation deviation from from the dummy element reference length. A trace is drawn for both the X and Y sensors while the length measurement meter reading is manually recorded on the form. A complete reading complete set of measurements measurements for all four elements in a bundle can be completed completed in about twenty minutes. UWNR Safety Analysis Report Rev. 2 9-10 9-10 Sept. 2008
*
- Operating Rod Extension Socket
'. Differential Transformers Clamping Device Figure 9-5 9-5 Measuring Tool Maintenance and Measuring Fuel Maintenance
- UWNR Safety Analysis Report Rev. 2 9-11 9-11 Sept. 2008 2008
*
- 0 Bow Sensors UWNR 9-6 Figure 9-6 UWNR Safety Bow and Elongation Elongat,ion Sensors Safety Analysis Report Report Rev. 2 9-12 Sept. 2008
*
- readouts are 900 Since the X and Y readouts 90° apart, the maximum maximum possible possible bow will be the square root of of indicated by the sum of the squares of the bows indicated by direct measurement.
measurement. As long as neither exceeds 0.088 measured bow exceeds 0.088 inch, no calculations calculations or other measurements measurements are necessary. necessary. If If either either measurement exceeds bow measurement 0.088 inch, then the square root of the sum of the squares exceeds 0.088 squares of the measured measured bows bows must be calculated calculated to determine determine whether whether or not this resultant 1/8 inch. resultant is less than 1/8 If calculated number If the calculated number is less than 1/8 1/8 inch, the element element is within technical specifications. specifications. Should the calculated calculated bow exceed 1/8 1/8 inch, the crowsfoot crowsfoot wrench wrench may be used to rotate the element element being being measured measured so that the reading reading of one bow sensor is maximized maximized and the true bow bow may be determined determined directly directly to see whether whether it exceeds technical technical specification specification limits. 9.3 9.3 Fire Protection Fire Protection Systems Systems andand Programs Programs detection and protection The fire detection protection systems systems in the laboratory meet meet the local and state requirements. requirements. Reactor Laboratory Interior doors from the Reactor Laboratory to the remaining remaining parts of the building building are also fire doors that meet local codes. All All walls between reactor laboratory between the reactor laboratory and the remainder remainder of of the building are masonry. An early early warning smoke detection detection system system within the laboratory is connected to the building fire alarm system.. The fire alarm connected alarm system alarms locally The laboratory laboratory is equipped equipped with a sprinkler sprinkler system system and portable portable fire extinguishers extinguishers which which are
- regularly regularly inspected and serviced by by the University Safety Safety Department.
Department. 9.4 9.4 Communication Com~unication Systems Systems The Reactor Laboratory and auxiliary support Reactor Laboratory support space space and offices are equipped with commercial and offices commercial telephones. telephones. Two lines are available at the reactor reactor control control center. A A cellular phone phone is also kept kept inin except when being the control room except being used for maintaining contact with the control maintaining contact control room. An intercom system is installed in the Reactor Reactor Laboratory Laboratory and auxiliary auxiliary support space and. and offices. This system provides provides two-way communications between two-way communications stations and an all-call between stations capability for paging. 9.5 9.5 Possession and Possession and Use Use of Byproduct, Source, of Byproduct, Source, and Special Nuclear and Special Nuclear Material Material All activities using radioactive All materials covered radioactive and special nuclear materials covered under the reactor license license place within the Mechanical take place Engineering Building Mechanical Engineering Building and a small portion of the Engineering Engineering Research Research Building Building in the rooms and areas indicated indicated below:
*
- Reactor Laboratory Laboratory (Room (Room 1215);
1215);
*
- Auxiliary Support Space Space (Rooms B 1135 and B Bl13S 1215, including B121S, sub-rooms);
including sub-rooms);
- Safety Analysis Report UWNR Safety Report Rev. 2 9-13 913Sp.20 9-13 2008 Sept. 2008
Radioactive and special nuclear material use outside these areas is conducted under Wisconsin Department Department of Health and Family Services license 25-1323-01, 25-1323-01, the University Wisconsin University of Wisconsin
- Radioactive Materials Materials License.
9.6 9.6 Cover Gas Control in closed Primary Coolant Systems There is no cover gas control in the primary coolant coolant system. 9.7 9.7 Other Auxiliary Systems There are no other auxiliary auxiliary systems required for safe reactor operation. 9.8 9.8 References References There are no references. 0
- UWNR UWNR Safety Analysis Report Rev. 2 9-14 9-14 Sept. 2008 2008
*
- 10 10 10.1 10.1 EXPERIMENTAL FACILITIES EXPERIMENTAL Summary Description Summary Description FACILITIES AND Facilities are provided to permit use of radiation Facilities endangering endangering personnel.
AND UTILIZATION UTILIZATION radiation from the reactor Facilities provided with this reactor personnel. Facilities experimental work reactor in experimental reactor include four beam work without without beam ports, a thermal column, column, and pneumatic and hydraulic hydraulic irradiation irradiation transfer systems. All All systems systems are designed to control control radiation radiation exposure exposure to personnel personnel using the the facility as well well as members members of the public. Consideration Consideration is given to controlling controlling the Ar-4l Ar-4 1 effluent from the experimental experimental facilities facilities in order order to meet both the limits on releases releases to the environment environment and personnel within and exposure of personnel within the laboratory. 10.2 10.2 Experimental Facilities Experimental Facilities 10.2.1 Thermal 10.2.1 Thermal Column Column The thermal column, Figure 10-1, 10-1, is a graphite-filled, graphite-filled, horizontal penetration through horizontal penetration through the biological biological shield shield which provides provides neutrons in the thermal energy energy range 0.025 eV) for range (about 0.025 irradiation experiments. The column, which is about 88 feet long, is filled with about irradiation experiments. about 6 6 feet of of graphite. graphite. A A small experimental experimental air chamber chamber between the face of the graphite and the thermal column connections (air, column door has conduits for service connections (air, water, electricity) electricity) to the biological shield biological shield face. Detectors Detectors for the safety safety channels channels and the LogN channel are located located within the thermal
- column. The location of the thermal column is indicated indicated in Figure Figure 10-2.
Personnel in the building protected against building are protected against gamma gamma radiation radiation from the column by by a dense concrete door which closes the column at the biological shield. The door moves on tracks set set into the concrete concrete floor perpendicular perpendicular to the shield face. A ventilation A system maintains a low pressure within the thermal ventilation system thermnal column column so that air flow is into the column when when the door is open. The door is gasketedgasketed so that air flow is very small when the door is closed. When the door is opened, opened, however, however, an air velocity velocity of about 40 feet per minute into the column column prevents Ar-4 1 activity from diffusing prevents the Ar-4l Reactor Laboratory. diffusing into the Reactor Laboratory. Section Section 9.1 contains 9.1 contains further information information on the ventilation ventilation system for the thermal columncolumn and beambeam ports. An annunciator annunciator is activated whenever whenever the thermal column column door is not not fully closed. In addition, an area radiation radiation monitor monitor beside beside the thermal column door will give an alarm should the reactor thermal column reactor be operated at a substantial substantial power power with the door open. open .
- UWNR Analysis Report Rev. 2 UWNR Safety Analysis 10-1 101Sp.28 lO-l 2008 Sept. 2008
- Q)
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*
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Figure 10-1 Figure 10-1 Thermal ThermalColumn Column UWNR Safety Analysis Report Rev. 0 UWNR Safety Analysis Report Rev. 0 10-2 10-2 April2000 April 2000 *
- 10.2.2 Beam Ports Four 6-inch beam ports penetrate penetrate the shield and provide provide fluxes of both fast and thermal neutrons experimental use. Figure 10-2 indicates the positions of the beam ports with respect to the for experimental grid box and shield while Figure 10-3 10-3 shows the construction of the beam ports. ports.
The ports are air-filled tubes, welded shut at the core ends and provided with water-tight water-tight covers on the outer ends. The portions of the ports within the pool are made of aluminum, while the portions within the shield are steel. A shutter assembly, made of lead encased in irradiations by a cable aluminum, is opened for irradiations lifting device that extends extends to the pool curb. When closed, the shutter shields against gamma rays from the shut-down core, allowing experiments experiments to be loaded and unloaded without excessive excessive radiation exposure to personnel. A drain line is attached to the bottom of the shutter housing, while a vent line beam of the top of attaches to the top the shutter housing. All shutter housing. All beam port drains combine before exiting the concrete
- shield, where a stop valve is provided.
provided . When beams of radiation are not being extracted, shielding plugs are installed in the outer end of of each port, filling almost all of the volume volume within. Figure 10-2 Location of Beam Ports & These plugs, made of dense concrete in aluminum Thermal Column Thermal Column casings, casings, have spiral conduits for passage of passage of instrument leads. These plugs completely completely stop the beam of radiation radiation and minimize the production of Ar-41 in the beam ports. Sealed aluminum aluminum cans are installed in the in-pool portion of the beam ports unless the particular particular beam port experiment experiment requires installation of a collimator collimator or filter in that location. These cans contain the Ar-41 produced and further minimize the release of Ar-41 activity. Additional control of Ar-41 activity released is accomplished accomplished by bellows seals on the lifting cables cables and maintaining the valve on the common drain header closed. Since extremely high radiation levels could exist should the reactorreactor be operated at substantial power levels with the shielding plugs removed, a beam port monitoring monitoring system is provided. The system consists of radiation detectors mounted on the walls in line with each beam port and a read-out device at the console which gives an audible and visual alarm should a preset radiation radiation level be exceeded. The system is set to alarm at a radiation level equivalent to a dose rate of of about 60 mremlhour mrem/hour at the beam port openings. openings .
- UWNR Safety UWNR Safety Analysis Report Report Rev. 2 10-3 10-3 Sept. 2008 2008
The beam port & & thermal thennal column ventilation system (Section through the vent pipes shown shown in Figure 10-3. through a check valve which prevents back-flow (Section 9.1) 9.1) exhausts the beam ports 10-3. Vent pipes are connected to the ventilation system back-flow into the vent and an isolation isolation valve which may be closed should the beam port fill with water. With the beam port open, a linear of about 40 feet per minute is maintained maintained into the port opening, preventing linear flow velocity preventing diffusion of the
- airborne activity into the laboratory. With the beam port closed Ar-41 is almost entirely entirely contained contained within the beam port.
- 0 UWNR Safety Analysis Report Rev. 00 10-4 10-4 April 2000 *
(i2 p.4 8
- Figure 10-3 Figure 10-3 Beam Beam Port Port UWNR UWNR Safety Safety Analysis Analysis Report Report Rev.
Rev. 00 10-5 10-5 April 2000 April 2000
10.2.3 Pneumatic I*1. 10.2.3 A Pneumatic Tube pneumatic tube is used to irradiate A pneumatic processed processed immediately radioisotopes. irradiate samples immediately after irradiation, radioisotopes. The The currently currently installed samples for a short time and irradiation, as in neutron act}vation installed pneumatic pneumatic tube and when when the sample sample must be act'ivation analysis of short-lived tube system system conveys conveys samples short-lived from the samples from the
- basement auxiliary support space to an irradiation position beside the core (Figure 10-4). The basement auxiliary support space to an irradiation position beside the core (Figure 10-4).
"rabbits" used "rabbits" used inin the system will the system convey samples will convey samples up to 1-1141-1/4 inches inches diameter and 5-112 5-1/2 inches inches long, although the gross weight weight of a sample sample is kept below 12 ounces. Although the polyethylene polyethylene rabbits used in the system can withstand longer irradiations, withstand longer irradiations, this facility is usually used for shorter irradiations irradiations of small objects. The system operates operates as a closed loop ioop with CO CO 22 cover gas controlling generation controlling generation and discharge of and discharge Ar-41 activity.
of Ar-41 activity. The system isispurged The system with CO purged with CO22 upon upon startup startup to remove any air which may have leaked in. An in-line in-line CO CO22 detector detector is used to monitor monitor Iconcentrations concentrations during the purge. Two isolation ball-valves isolation ball-valves are installed just outside outside of the Ibiological biological shield which are closed closed unless the system is running. These valves valves serve as a barrier barrier Iagainst against air leaking into the reactor portion reactor portion of piping when not in use, as well as preventing loss of preventing of pool water if the internal internal piping should rupture. IThe The reactor control room pneumatic system panel reactor control panel includes pneumatic pneumatic system "System "System Start" and I"System "System Stop" Stop" buttons, an "Emergency "Emergency Return"Return" button (which(which allows the reactor reactor operator operator to Ireturn return rabbits to the fume hood hood if desired), indicator lights desired), and indicator lights for "System "System On," "System "System Purge Progress" (for CO In Progress" CO 22 purge), "Reactor "Reactor Isolation Valves Valves Open," Open," and "Rabbit Reactor."* All "Rabbit In Reactor." All Iindications indications and controls except for the system system start capability capability are duplicated duplicated at the pneumatic pneumatic
- Itube tube control center in the basement control center basement auxiliary auxiliary support space immediately support space immediately west of the ReactorReactor Laboratory.
Laboratory. Automatic Automatic timing of irradiations irradiations is done at this controlcontrol center, and rabbits are inserted, dispatched, and removedremoved at this location. location. An area monitormonitor indicates indicates radiation level and gives a visible and audible alarm should the radiation level exceed exceed a preset level at the station. The preset preset level is selected according according to the computed computed activity of the sample irradiated. sample being irradiated. IThe The pneumatic station is installed pneumatic tube station installed in a fume hood with a high efficiency filter to control control any releases releases from sample failures. Sample Sample activity is limited to a level which, should the sample sample rupture upon discharge discharge from the system, system, will result in keeping the concentration concentration exhausted exhausted below below I10 10 CFR Part 20 limits for unrestricted unrestricted areas when averaged over averaged over perioda period of 24 hours for routine samples, or 30 30 days for non-routine samples (requiring RSC approval). non-routine samples approval). The reactivity reactivity effect from a sample is restrictedrestricted to less than 0.2% 0.2% p.p. Tests run with waterwater and and cadmium cadmium samples indicate that sample reactivity effects will normally reactivity effects normally be lessless than 0.0 0.011%
% p.p.
reactivity measurements Static reactivity measurements will be run for samples samples of fissionable fissionable material or particularly particularly strong absorbers absorbers such as some of the rare earths. Since Since the pneumatic pneumatic tube penetrates penetrates the shield shield below water level, a leak leak in the tubing tubing could draindrain Ithe the pool. Spring-close Spring-close air-open air-open automatic automatic ball valves located in the tubing just outside the shield shield Iautomatically automatically close when the pneumatic pneumatic system is turned turned off. Should Should these valves fail to close, close, water water will not be lost from the system from a break inside the pool unless another break in the will not be lost from the system from a break inside the pool unless another break in
- the UWNR U"R Safety Analysis Safety Analysis Report Rev. 2 10-6 10-6 2008 Sept. 2008
- tubing occurs outside the pool. Further, to drain more than 8 feet of water from the pool a siphon action would have to be set up. A siphon action is prevented prevented by a solenoid solenoid valve controlled siphon breaker breaker at the highest point in the system. The solenoid valves close when the pneumatic system is started. When the pneumatic system is off,off, the solenoid valves open and check check valves will then allow air to enter the system if a siphon action starts. Normally these check valves prevent loss of cover cover gas from the system.
The system is operated using a written check-list type procedure to assure that the built-in safe-safe-guards remain effective.. remain effective
- Figure 10-4 Pneumatic Tube System Layout Pneumatic
- UWNR Safety Analysis Report Rev. 2 10-7 10-7 2008 Sept. 2008
10.2.4 Grid Box Irradiation 10.2.4 Irradiation Facilities Irradiation of larger larger samples and most irradiations of more than twenty minutes duration performed in irradiation irradiation facilities on the core periphery inside the grid box. duration are
- Radiation baskets (Figure 10-5)10-5) are 3-inch square aluminum containers containers which fit into the grid plate and may contain contain one or more samples. The bottom bottom end boxes are similar similar to those of of reflector elements, thus positioning positioning the devices in fixed positions relative to the core. These devices may contain internal shelves or other positioning devicesdevices to position samples in fixed positions.
Figure 10-6 shows a hydraulic irradiation tube (called Figure 10-6 (called a whale tube at the facility). In this facility sample movement is powered by a separate separate pump located beneath the north side of the reactor reactor bridge. The bottom ends of these tubes fit into the grid plate, and the top of the tube is fastened to the bridge structure structure to provide further support and prevent prevent inadvertent inadvertent movement. The motor for the pump is electrically electrically paralleled paralleled to the diffuser pump and thus runs when the diffuser pump is in operation. The pump takes its suction just below the pool surface surface and directs a jet pump near the bottom end of each tube, causing its flow to ajet causing sufficient induced flow down the tube to move samples to the irradiation irradiation position and hold them in place. Samples which float return to the top of the tube where they are retained until removal by operating personnel. Non-floating samples can be removed with a retriever tool, or they may be installed with a retrieving string or wire attached. Flow direction and "sample in" indicators and controls are located at the
- pool top and control console.
Rupture of piping connected connected to the hydraulic irradiation facility will not result in loss of pool water due to its location within and immediately immediately above the pool. Reactivity Reactivity effects of samples samples are much smaller than those associated with installation and removal of conventional conventional irradiation irradiation baskets with samples in them. The remarks regarding reactivity effects effects for samples in the pneumatic pneumatic tube apply to this facility. UWNR Safety Analysis Analysis Report Rev. 2 10-8 10-8 2008 Sept. 2008
*
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- 10.3 10.3 Experiment Review Experiment A
A body of operating procedures is in place place to assure that experiments experiments are conducted conducted in a manner manner that will ensure the protection of the public. Experiment Experiment review review meets the requirements requirements of of Regulatory Regulatory Guide N401-1974/ANS-15.6 22 as modified by Regulatory Guide 2.21 and standard ANSI N401-1974/ANS-15.6 Guide 2.4 3.'. UWNR 002, UWNR 002, Experiment Experiment Standing Operating Operating Instructions, defines several classesclasses of experiments experiments that are routinely conducted and states the limitations limitations and precautions to be observed as well as the methodology methodology to be used. Control Control Element Calibrations, Calibrations, reactivity coefficient coefficient measurements, in-core neutron flux distribution distribution measurements measurements and sample sample irradiation/isotope production irradiation/isotope production experiments experiments are specifically specifically defined in these instructions. Since sample irradiation and isotope production production are major experimental experimental activities at UWNR, several additional standard standard operating procedures and limitations are in effect for these activities. Limits on potential potential airborne radioactivity produced produced in the event event of sample breakage are included sample breakage in the procedures procedures to assure releases releases will not exceed exceed those considered considered in the safety safety analysis report or those permitted permitted under technical specifications. irradiation Irradiation of fueled experiments experiments is controlled controlled so that the total inventory inventory of iodine isotopes 131 through 135 in the experiment experiment is no greater than 1.5 Curies. UWNR 002 allows SRO approval of irradiations meeting the requirements of of UWNR UWNR 131 up to the limits for routine approval of gas, dust, highly volatile material, and and
- fissionable material stated on UWNR 130, Request For Isotope Production. Approval Approval for limits above those stated in UWNR UWNR 130 but below below 10 CFR Part Part 20 limits when averaged over 30 days of dilution require approval approval by the Reactor Safety Committee. Other Other written written procedures procedures in the UWNR 130 series, including including sample packaging requirements requirements for the different irradiation irradiation facilities and approvals, are in effect for operation operation of all experimental experimental facilities. Irradiation of Irradiation of material that is to be transferred transferred to the campus broad radioisotope license requires both written writ~en and telephone telephone approvals to assure that the recipient of the material is permitted possession and use of the material material under that license.
For other experiments experiments the senior reactor operator (SRO) responsible operator (SRO) responsible for operation when the experiment is performed experiment performed classifies the experiment experiment as routine (previously approved and performed), modified routine (determined not to be significantly different from previously performed), performed experiment), special (not previously performed experiment), previously approved, approved, but within constraints of technical specifications), specifications), or special requiring NRC approval (involving (involving technical specification specification changes changes or unreviewed questions). . unreviewed safety questions) Routine Routine experiments may be approved approved by the SRO without further evaluation. For other experiments, the SRO evaluates the experiment experiment in terms of its effect on reactor operation and the possibility possibility and consequences experiment failure, including consequences of experiment including consideration consideration of chemical reactions, reactions, physical integrity, design life, proper cooling, reactivity effects, effects, and interaction interaction with core components. components. IftheIf the experiment experiment is classified as modified routine then two SROs may approve operation operation of the experiment experiment if it is determined and specified specified in a written record record that the
- UWNR Safety Analysis Analysis Report Report Rev. 2 10-11 10-11 2008 Sept. 2008
experiment are not significantly hazards associated with the modified routine experiment significantly greater or different different
- involved with t~e from those involved the corresponding corresponding routine experiment.
experiment is determined to be a special experiment, an experiment review questionnaire If the experiment (UWNR 030) which includes description of the experiment includes description experiment including materials inserted, thermodynamics, reactivity radioactivity, shielding, instrumentation related to control reactivity effects, radioactivity, procedures, as well as a safety analysis must be completed administration, and procedures, panel instruments, administration, completed and reviewed. Special experiments must be reviewed and approved by the Reactor Reactor Director and evaluation of an experiment shall conclude that failure the Reactor Safety Committee. Favorable evaluation reactor fuel or interference experiment will not lead directly to damage of reactor of the experiment ofthe interference with movement of a control element. Special requiring NRC approval approval experiments experiments require require local approvals as well as approval by NRC. 10.4 References References Development of Technical
- 1. Regulatory Guide 2.2, Development
- 1. Specifications for Experiments Technical Specifications Experiments in Research Research Reactors, US Nuclear Reactors, Regulatory Commission, November 1973 Nuclear Regulatory
- 2. American National Standard Standard ANSI N401-1974/ANS Experiments for N401-1974/ANS 15-6, Review of Experiments Research Reactors, American Nuclear Nuclear Society, November 19, 1974 1974
- 3. Regulatory Guide 2.4, Review of Experiments for Research Regulatory Commission, May 1977 Regulatory Reactors, U. S. Nuclear Research Reactors, Nuclear
- UWNR Safety Analysis Report Rev. 2 10-12 10-12 2008 Sept. 2008 *
- 11 11 RADIATION PROTECTION RADIATION This chapter PROTECTION PROGRAM PROGRAM AND chapter deals with the program and procedures the University of Wisconsin-Madison AND WASTE MANAGEMENT MANAGEMENT procedures for dealing with radioactive materials, radiation, and radioactive waste management. Since the Nuclear Laboratory is a part of Nuclear Reactor Laboratory Wisconsin-Madison the campus radiation safety regulations govern activity under the reactor license. Information on these campus regulations was promulgated of promulgated under Wisconsin Department of Health and Family Services Services license license 25-1323-01. The Radiation Safety 25-1323-01.
i Regulations' Regulations are also available available on the UW Radiation Radiation Safety information is Safety website. This information incorporated by reference as part of this Safety Safety Analysis Report. The intent of the campus radiation safety programprogram is to maintain radiation exposure exposure to experimenters, students, and the general public as low as reasonablyreasonably achievable as well as below below regulatory limits while using radiation and radioactivity radioactivity for teaching and research purposes. The implementation of the campus implementation campus program within the activities of the Nuclear Reactor Reactor Laboratory has the same intent. 11.1 11.1 Radiation Protection Protection The radiation protection protection program at the reactor facility, while conforming conforming to the campus program, has some specific aspects that apply only to the reactor reactor facility. For instance, the design of the experimental experimental facilities, the reactor pool, and the reactor reactor shield includes protective protective
- measures measures and devices which limit radiation exposures and release of radioactive radiation exposures radioactive material to the environment. Information on these aspects of the radiation environment. Information radiation control program is included in the sections sections of this report that describe describe that equipment. General requirements, such as dosimeter dosimeter use and records, certification certification of training, survey frequency, leak testing of sources, and overall ALARA ALARA program program are discussed in the campus documentation.
documentation. The remaining portions of this chapter chapter will deal with the issues specific to the reactor. Sources 11.1.1 Radiation Sources 11.1.1.1 Airborne Radiation Sources Radiation Sources 11.1.1.1.1 Releases from abnormal abnormal reactor reactor operations The fuel retains the fission products, with releases to the environment environment only if the fuel clad is breached. This possibility is one of the accidents considered considered in Chapter Chapter 13 of this report report in the analysis of the maximum hypothetical hypothetical accident. This event would result in maximum dose to personnel within the Reactor Laboratory and the maximum dose released to the environment. Reactor Laboratory The maximum occupational occupational dose calculated calculated is 10 millirem whole body, 1 rad to the lung, and 18.9 rads to the thyroid, while the maximum dose to persons in unrestricted unrestricted areas will be less than 0.153 rem whole body and 1.019 1.019 rad to the thyroid. thyroid .
- UWNR UWNR Safety Analysis Report Rev. 2 11-1 11-1 Sept. 2008
11.1.1.1.2 11.1.1.1.2 Calculations Releases from normal reactor Releases Argon-41 is the only activity released reactor operations released in significant quantities Calculations and measurements have been performed to determine of the various activities that might be discharged quantities during normal operations. production and release rates determine production discharged due to normal operation. The calculationcalculation rates
- method used for Ar-41 release is shown shown in Appendix A, SectionsSections A and B.
Due to the operation operation of the beam beam port and thermal column column ventilating ventilating system and the laboratory exhaust fan, the airborne exhaust airborne activity levels in the laboratory are low. Some Ar-41 is produced in the dissolved air in the pool water as it passes passes through the reactor core and is released as the water water is warmed warmed while passing through the core. Some of the resulting activity eventually eventually reaches the pool surface where it is released released to the laboratory atmosphere. The concentration concentration of Ar-41 in in the air immediately immediately above the pool surface during full-power operation operation reaches about one-third of the DAC for occupational occupational exposure; exposure; as this air diffuses throughout throughout the laboratory, the activity laboratory as a whole is at least a factor of 6 below the DAC. Therefore, further discussion in the laboratory discussion will be concerned with the activity activity released released to the atmosphere. The maximum release release rate of Ar-41 would occur with the reactor reactor operating continuously at 1I operating continuously MW and all four beam ports and the thermal column open. Such Such operation is not reasonable, reasonable, but it does establish an upper limit to the activity that might be discharged. This maximum release rate is 13.3 !lCi/sec, iCi/sec, giving an Ar-41 concentration concentration at the stack of 2.94xl 0-66 !lCi/ml. stack outlet of2.94x10- [tCi/ml. The EPA COMPLY program program22 indicates that the maximally exposed exposed receptor would receive a dose of of
- 0.6 mremlyear mrem/year if all activity generated activity generated were discharged continuously.
discharged continuously. The maximum concentration concentration to which the public would be exposed (using Gifford's model as discussed in Appendix discussed Appendix A and assuming assuming a zero stack 3.31 xl -9 stack height) in this case would be about 3.31x10-9 [tCi/ml. !lCi/ml. As previously previously indicated, the above maximum value is for a situation situation not likely to occur during during operation. The usual procedure procedure is to have the experimental experimental facilities in a no-flow condition if condition if possible. Under Under no-flow conditions the beam port and thermal column ventilation system keeps column ventilation the pressure in the experimental experimental facilities lower than room pressure, and the activity activity produced produced in the facilities remains there and decays. The ALARA ALARA measures measures taken on the experimental experimental facilities limits the typical release release rate to about 10% 10% of the production rate. Historically, in the year in which the maximum maximum recorded Ar-41 release to the environment occurred (1999-2000 (1999-2000 fiscal year), the COMPLY program indicated a resulting dose of 0.004 mrem/year. mremlyear. One theoretically theoretically important consideration in the analysis of a reactor location is the effect on important consideration surrounding unrestricted surrounding unrestricted areas of a spillage of radioactive radioactive materials. A releaserelease of radioactive material material might occur, for example, if a highly-volatile highly-volatile liquid were irradiated irradiated in the reactor for the production of isotopes. If, If, while it was being transferred transferred from the reactor to a cask, it were dropped and its container broken, the atmosphere atmosphere within within the Reactor Laboratory Laboratory could become UWNR Safety Analysis Report Rev. 2 11-2 11-2 2008 Sept. 2008
*
- conceivably contaminated; further, this atmosphere conceivably contaminated; atmosphere could conceivably be released could conceivably released to the surroundings in such a fashion as to present present an exposure exposure in unrestricted unrestricted areas.
For a typical solid or liquid spill, no special problems problems exist other than the direct radiation from the sample and cleaning up contamination. contamination. Since the level level of radiation radiation will be known for each each sample, adequate equipment for handling the sample will be available when the material is material is discharged from the reactor. Equipment discharged Equipment adequate for cleanup cleanup of spills will be kept available available soso that spills can be dealt with immediately, immediately, lessening the possibility of spreading contamination contamination to adjacent areas. adjacent The remainder of this section section will deal with gases, highly volatile liquids, or powdered powdered samples samples which might cause air-borne air-borne activity in the event of a spill. This problem is handled at Wisconsin by a combination combination of administrative administrative and operational operational procedures. procedures. For routine operations, a concerted concerted effort will be be made to keep the concentration contaminants in the concentration of contaminants atmosphere released from the Reactor Laboratory Laboratory well below below the limits as stated in Table Table 2, Appendix B, 10 CFR Part 20, "Standards for Protection Protection Against Radiation." Among Among the procedures which which will be followed to achieve this goal will be the double-encapsulating double-encapsulating of of materials to be exposed exposed in the reactor in aluminum aluminum containers containers (for long exposure) sealed exposure) or sealed polyethylene containers of less than 4 x 10"7 containers for exposures ofless 17 10 thermal neutrons/cm 2 neutrons/cm with accompanying gamma ray and fast neutron fluxes. Only members of the reactor accompanying selected reactor staff (or selected and trained individuals working under their supervision) supervision) will be permitted to handle these capsules within the Reactor Reactor Laboratory Laboratory and the capsules will normally normally be opened opened only at
- appropriate appropriate locations outside material generated in any radioactivity outside the laboratory. Further, a log book will be maintained material exposures. However, because accidents will be generated anyone accidents can occur, the amount one sample of material maintained of all amount of radioactivity radioactivity which material will be limited. Specifically, this amount of radioactivity will be limited for routine samples such that, should a container be broken and its contents contents disperse disperse in the air within the Reactor Laboratory, the concentrations concentrations discharged of discharged through the stack when averaged over 24 hours will not exceed concentrations concentrations of 10 CFR Part 20 Appendix B Table 2. In normal operation (single (single fan running), the ventilation system has a capacity capacity of 9,600 scfm through through its filters, therefore 24 hours of dilution is 3.9x10"3.9xl 011 ml. For non-routine samples, RSC approval approval will be required, but the above limits will still apply assuming a 30 day dilution instead of 24 hours. 30 days of dilution is 1.2x 1.2x10i3 10 13 ml. These approvals will consider all other activity activity discharged, and will insure that the total stack discharge lies within permissible limits should the sample rupture.
The pneumatic pneumatic tube station is located in the Reactor Laboratory Laboratory auxiliary supportsupport space space and thus it is subject to the laboratory laboratory ventilation system. The station is installed within a fume hood hood having a face velocity of of> > 100 lfpm to protect the system operator in case of sample breakage. breakage. The air discharged discharged from the hood is passed through a dedicated high efficiency efficiency filter before connecting to the main laboratory laboratory ventilation exhaust. packaging requirements Although special packaging enforced to prevent breakage of pneumatic tube requirements are enforced samples, samples, such breakage breakage may occur. SampleSample activity is limited as discussed above, assuming assuming 24
- UWNR UWNR Safety Analysis Report Report Rev. 2 11-3 11-3 2008 Sept. 2008
hours of normal ventilation ventilation dilution for routine samples, samples, and 30 days of dilution for non-routine
- requiring RSC approval. The fume-hood blower samples requiring blower operates automatically automatically whenever whenever the pneumatic tube system is used. As with the other samples, the maximum activities generated generated for for non-routine samples must have RSC approval, and only quantities considerably considerably smaller are routinely approved.
11.1.1.2 Liquid Radioactive 11.1.1.2 Sources Radioactive Sources The only activity produced produced in liquid form in amounts sufficient sufficient to be a personnel exposure hazard is Nitrogen-16, Nitrogen- 16, which is produced produced in the reactor coolant as it passes through the reactor reactor core when operating core operating at power levels above 100kW. controlled by use of the diffuser 100kW. N-16 is controlled system (discussed (discussed in Section Section 5.6), which reduces the dose rate at the pool surface to 22 to 3 mrem/hour mremlhour during full power operation. If the diffuser system fails during full power power operation operation the dose rate at the pool surface surface is less than 100 mremlhour. Small quantities of liquid radioactive waste are generated generated by regeneration of the demineralizer demineralizer and from liquids irradiated as part of sample irradiation. The radiation level from such liquids is extremely low and does not produce produce radiation exposure exposure hazards. Disposal of this material material is addressed in section 11.2.3. 11.2.3. Releases are made to the sewer system within 10 CFR Part 20 20 Appendix B TableTable 3 3 limits. Annual liquid releases releases have ranged from 0 to 10,000 gallons, with 3000 gallons being typical. 11.1.1.3 Solid Radioactive Sources 11.1.1.3 fuel. Typical four-element air at 3 feet if Sources The major source of radiation and radioactivity radioactivity is the fission product generation if removed from the reactor reactor pool. generation in the reactor four-element fuel bundles will generate fields of 100 to more than 1000 Rlhour R/hour in
- As long as the fuel is contained contained within the pool filled with water water this source source of radiation dose presents presents no personnel personnel hazard. Loss of pool water is considered in considered in Chapter 13,13, with the conclusion conclusion that the dose rates from pool water loss after long periods of of operation operation could result in high radiation levels at the pool top (1200 R/hour one day after (1200 Rlhour shutdown), but not so high that persons could not perform perform corrective corrective actions to restore enough pool level to reduce the dose rate to tolerable levels (dose rate at the pool top level when shielded shielded by the pool curb would be about 240 mrem/hour mrem/hour at the same decay time). Further, the pool is designed to preclude preclude loss of pool water, and operation would not take place if there were any difficulty in maintaining maintaining pool level.
Other possibilities of significant radiation exposure from solid radioactive radioactive material material are the standard 20% 20% enriched TRIGA core, core, samples irradiated irradiated for isotope production, reactor reactor components which have spent a long time near the core, and the reactor components reactor startup source. All of of these are small sources sources compared compared to fuel fission product activity in the operating operating core. Dose UWNR Safety Analysis Report Rev. 2 UWNR 11-4 11-4 2008 Sept. 2008
*
- rates from the old fuel are several orders orders of magnitude magnitude lower than those from from the operating core.
Sample handling equipment and procedures procedures and use of aluminum for almost all structure structure near the core reduce exposure exposure rates from samples and activated materials to levels which generate generate no significant personnel hazard during operation operation or maintenance maintenance of the reactor. For example,example, the shim-safety blades, reflector elements, and transient control rod have maximum radiation levels levels R/hour at contact after a week of reactor shutdown. Activity produced during of a few Rlhour irradiations is calculated before before the irradiations irradiations are performed performed and equipment equipment and procedures procedures are in place to deal with the activity after the irradiation irradiation is completed. 11.1.2 Radiation Protection Program Program 11.1.3 ALARA Program 11.1.3 Program Note: These two sections sections are combined. The University Radiation Regulations!1 are written to incorporate Radiation Safety Regulations incorporate ALARA ALARA principles and practices. practices. The Nuclear Nuclear Reactor Laboratory policies and proceduresprocedures reflect the commitment commitment to ALARA principles. ALARA principles. An annual ALARA ALARA review is conducted conducted jointly by campus Safety Department health physics staff and the ReactorReactor Laboratory staff staff with a report of the results of of the review being submitted to the Reactor Director and the Reactor Safety Safety Committee. 11.1.4 Radiation Monitoring and Surveying
- The campus regUlations!1 specify campus regulations reactor operation reflect reflect specify requirements these requirements on monitoring and surveying. Procedures requirements. Installed described in Section 7.7 of this report. Area radiation including checks checks for and forms require checks contamination contamination and radiation radiation surveys particulate particulate air and air Procedures for activity monitors are surveys are conducted each month, activity. Sample irradiation irradiation procedures checks of radiation level each time a sample is removed from an irradiation procedures irradiation facility. Experiment Experiment reviews and approvals approvals require radiation surveys surveys for new experiments experiments and modifications of experiments.
11.1.5 Radiation 11.1.5 Radiation Exposure Exposure Control and Dosimetry The campus regulations! regulations' specify requirements requirements on radiation radiation control and dosimetry, and the Safety Department Department administers administers the dosimetry dosimetry program. TLD dosimeters dosimeters are used for operating personnel personnel and experimenters experimenters using the laboratory laboratory on a regular basis, and electronic dosimeters are used and records records are maintained maintained for tour groups and visitors. Experiment approval requires that no Very Experiment Radiation Areas are created external Very High Radiation external to the experiment shielding. Some experiments experiments have shield cavities cavities large enough enough for personnel entry, however, and higher radiation levels can exist inside the shield. Should Should an experiment experiment design be approved with a Very High Radiation Level within the experiment experiment shield, protective measures protective measures will be in place that will reduce radiation radiation levels to no more than a high radiation area if access is
- UWNR Analysis Report UWNR Safety Analysis Report Rev. 2 11-5 11-5 Sept. 2008 2008
Area is created inside or outside of the shielding, access will be attempted. If a High Radiation Area
- 20.1601.
controlled and posted in accordance with 10 CFR 20.1601. controlled Radiation doses received by visitors and tour groups are so low that they routinely cannot be measured; measured; the maximum mrem/hour and for any non-maximum dose rate allowed for any tours is 0.5 mremlhour mrem/hour. Visitors, who are radiation radiation workers is 2.0 mremlhour. radiation workers not part of the campus researchers, are allowed higher dose rates, but rarely do they dosimetry program such as visiting researchers, mrem/hour due to ALARA practices. No student dosimeter has ever received exceed 2.0 mremlhour received a measurable exposure Occupational exposures exposure from reactor operation. Occupational exposures of operations and maintenance personnel have historically maintenance exceeding 0.5 Rem TEDE in a historically been very low, seldom exceeding year and usually below 100100 mrem/year. 11.1.6 Contamination Contamination Control regulations11 specify requirements The campus regulations Contamination Control. As noted in section requirements on Contamination section 11.1.4, monthly contamination 11.1.4, Laboratory policy is that no detectable contamination surveys are conducted. Laboratory contamination discovered removable contamination is allowed; any contamination discovered is immediately immediately decontaminated. decontaminated. laboratory has cleaning equipment For routine cleaning, the laboratory equipment which is dedicated dedicated to use in the laboratory area, and custodial personnel use this equipment in order to prevent the possibility laboratory of possibility of unidentified contamination. spreading unidentified contamination. Floor sweepings are surveyed surveyed for radioactivity radioactivity before
- disposal.
11.1.7 Environmental Environmental Monitoring Monitoring Environmental TLD monitors are used and evaluated on a quarterly basis. The dosimeters are Environmental distributed around the engineering engineering campus surround the Reactor Laboratory. At the campus so that they surround present time more than 25 points are monitored. Effluent concentrations are measured Effluent concentrations measured at the point of release. 11.2 Management Radioactive Waste Management regulations11 specify requirements The campus regulations radioactive waste on campus. requirements for dealing with radioactive The Reactor Reactor Laboratory follows the campus campus regulations. 11.2.1 Radioactive 11.2.1 Management Program Radioactive Waste Management This is a campus program. UWNR UWNR Safety Analysis Report Rev. 2 11-6 11-6 2008 Sept. 2008
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- 11.2.2 Radioactive Waste Waste Control This is a campus-wide program. Liquid waste from beam port drains, pool overflow, laboratory floor drains, radioactive sink, and demineralizer regeneration is stored in a 2000-gallon holdup demineralizer regeneration tank (see Chapter 5, 5, section section 5.5),
5.5), and other liquid radioactive wastes generated in the laboratory are collected in local containers. containers. Filled local containers containers may be dumped into the holdup tank. 11.2.3 Release of Radioactive Waste 11.2.3 Solid radioactive radioactive waste is transferred to the Safety Department Department for disposal. wastes can be transferred Liquid wastes Department, but most are placed into the holdup transferred to the Safety Department, tank. The Reactor Reactor Laboratory discharges liquid waste from the holdup tank to the Laboratory occasionally discharges sewer system. All discharges discharges are filtered so particulate activity above 0.5 micron size is that no particulate is discharged. Sampling, analysis, and release release of the holdup tank contents are governed governed by a W1,"itten written procedure releases are within 10 CFR Part 20 Appendix B Table 3 limits and procedure that assures releases within local limits for discharge that the pH is within discharge to the sewer. 11.3 11.3 References References
- 1. Radiation Wisconsin-Madison, Revision 2, Radiation Safety Regulations, University of Wisconsin-Madison, 2, January 1997.
- 2. U. S. Environmental Environmental Protection Agency, COMPLY Program Program Rev. 2, October 19891989
- UWNR Safety Analysis UWNR Analysis Report Report Rev. 2 11-7 11-7 Sept. 2008 2008
- This page is intentionally intentionally left blank.
- UWNR Safety Analysis UWNR Safety Analysis Report Report Rev. 2 11-8 11-8 Sept. 2008 2008
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- 12 12 CONDUCT OF CONDUCT OF OPERATIONS OPERATIONS 12.1 12.1 Organization Organization 12.1.1 Structure 12.1.1 Structure Figure 12-1 is a chart indicating Figure 12-1 indicating the operating operating organization.
organization. Position Position responsibilities responsibilities and and authorities authorities are summarized summarized in the following sections. sections. Responsibility 12.1.2 Responsibility 12.1.2.1 12.1.2.l University Radiation Safety University Safety Committee Committee 1.
- 1. To exercise its prerogatives prerogatives (as a campus-wide campus-wide committee appointed by committee appointed by the Chancellor Chancellor of the University Wisconsin-Madison Campus to review all University of Wisconsin-Madison activities activities on campus which which involve the use of radiation) radiation) in reviewing reviewing all activities activities related to the Reactor Reactor Laboratory.
Laboratory.
- 2. To advise the Reactor Reactor Director Director of all studies and/or actions taken with regard regard to the Reactor Reactor Laboratory.
Laboratory. 33..
- overrule the Reactor To overrule Reactor Director where carrying out where necessary in carrying out its function.
- 4. To supply services to the University.
supply health physics services University. 12.1.2.2 12.1.2.2 University Radiation Safety University Safety (Part of University Department of Environment, University Department Environment, Health and Safety) 1.
- 1. To assist the University University Radiation Radiation Safety Committee by Safety Committee conducting inspections, by conducting inspections, recommendations, maintaining making recommendations, maintaining records, and establishing procedures for establishing procedures emergency operations, emergency operations, waste disposal, disposal, etc.
- 2. To provide provide similar similar inspections and service functions to the Reactor Reactor Safety Committee.
Committee. 12.1.2.3 12.1.2.3 Chair, Engineering Engineering Physics Department Physics Department 11.. Responsible for the reactor facilities licenses Responsible licenses and charter. charter. 2.
- 2. To appoint the Reactor Director.
- UWNR UWNR Safety Safety Analysis Report Rev. 22 121Sp.08 12-1 2008 Sept. 2008
12.1.2.4 12.1.2.4 1. 1. Reactor Director Reactor Director To approve all policy decisions basic procedures decisions and all basic regulations, procedures governing facilities. governing the use and regulations, basic instructions, operation of the reactor and operation instructions, and reactor and related related
*
- 2. "To designate
!To designate the Reactor Reactor Supervisor Supervisor and other Senior operators.
other Senior operators. 3.
- 3. To take cognizance cognizance of all recommendations recommendations and actions by the University actions by Safety Committee Radiation Safety Committee (which relate relate to the reactor facility) and the Reactor Reactor Safety Committee.
Committee.
- 4. members to the Reactor To appoint qualified members Reactor Safety Safety Committee Committee as necessary.
necessary. 12.1.2.5 12.1.2.5 Reactor Safety Committee Reactor Safety Committee 1.
- 1. Review and approval Review experiments utilizing the reactor facilities; approval of new experiments facilities;
- 2. Review Review and approval of all proposed proposed changes changes to the facility, procedures, procedures, license, license, and technical technical specifications; specifications;
- 3.
- 3. Determination Determination of whether a proposed experiment would proposed change, test or experiment constitute would constitute an unreviewed safety question unreviewed safety question or a change in Technical Technical Specifications; Specifications;
- 4. Review of abnormal abnormal performance performance of plant equipment and operating plant equipment operating anomalies anomalies having safety significance; significance; and
- 5. Review Review of unusual unusual or reportable occurrences and incidents reportable occurrences incidents which are reportable reportable under 1010 CFR Part 20 and 10 10 CFR Part 550.
0. 6.
- 6. Review of audit reports.
reports. 7.
- 7. Review of violations of technical Review technical specifications, specifications, license, or procedures procedures and orders significance.
having safety significance. 12.1.2.6 12.1.2.6 Reactor Supervisor Reactor Supervisor 1.
- 1. enforce policies, administrative To initiate and enforce administrative rules, regulations, and operating operating procedures relating to the Reactor procedures Reactor Laboratory, subject to the appropriate appropriate approvals approvals of the Reactor Reactor Safety Committee, Committee, the University University Radiation Radiation Safety Safety Committee, Committee, and Reactor Director.
the Reactor Director. UWNR Analysis Report Rev. 2 UWNR Safety Analysis 122Sp.08 12-2 12-2 2008 Sept. 2008
- 40
- 2.
- 2. To ensure that all activities activities within the Reactor Laboratory Laboratory are in accordance accordance with prior approvals appropriate committees approvals from the appropriate committees or from the Reactor Director.
3.
- 3. The Reactor Supervisor shall have authority to authorize Reactor Supervisor authorize experiments and/or procedures which have been approved by the Reactor Safety Committee. He will procedures prepare specific detailed procedures based based on the general procedures procedures approved by the Committee. .
- 4. To see that all proper records are kept.
5.
- 5. To maintain a Senior Operator's Operator's License.
- 6. To appoint Reactor Operators.
Operators.
- 7. The Reactor Reactor Supervisor or another Senior operator shall be in charge of the Laboratory at all times (although Reactor Laboratory (although not necessarily necessarily physically present). The individual in charge, if physically present, shall be responsible for prompt prompt execution of emergency procedures. The Reactor Reactor Supervisor or another Senior Senior operator will be present at the facility during fuel manipulation, reactor start-up and approach approach to power, and recovery from unscheduled unscheduled scrams and shut-downs, shut-downs, and shall be available available on call at other times during reactor operation.
operation.
- 8.
9. To be responsible for safety in the Reactor Laboratory, including responsibility health physics physics matters. To advise and prepare Laboratory, prepare information information for the committees Laboratory, and to present such information responsibility for committees concerned with the Reactor information to the committees Reactor 12.1.2.7 12.1.2.7 Operators (alternate Senior Operators (alternate Supervisors) Supervisors) 1.
- 1. To accept responsibility for safe and efficient efficient operation of the Reactor Reactor Laboratory when when designated by the Reactor Supervisor.
- 2. To maintain a Senior Operator's License.
12.1.2.8 Reactor operators 1.
- 1. To hold a Reactor Reactor operator's operator's License.
- 2. To conform to all rules and regulations regulations for operation operation of the reactor.
- 3. A reactor operator operator will be present at the control console at all times when the reactor reactor is in operation.
operation .
- UWNR UWNR Safety Safety Analysis Report Report Rev. 2 12-3 12-3 Sept. 2008 2008
4. 12.1.3 Staffing 12.1.3 To monitor laboratory laboratoryactivities activities from a health-physics The minimum staffing when the reactor is not secured shall be: health-physics standpoint.
- 1.
- 1. A licensed licensed reactor reactor operator operator in the control room (if senior operator licensed, may also be the person required in 3 below)~
below):
- 2. A second designated personperson present present at the facility complex able to carry out prescribed prescribed written instructions.
3.
- 3. A designated designated senior reactor operator shall be readily available available at the facility or on on call.
A list of reactor facility personnel personnel by name and telephone telephone number shall be readily available in the control room for use by the operator. A licensed senior reactor reactor operator shall be present at the facility for: 1.
- 1. Initial startup and approach to power.
2. 3. All fuel handling Relocation handling or control-element control-element manual manipulations. Relocation of any in-core AK/K. LlKlK.
,
manipulations. experiment with a reactivity worth greater in-core experiment greater than 0.7% *
- 4. . Recovery Recovery from unplanned unplanned or unscheduled shutdown or significant significant power power reduction.
12.1.4 Selection 12.1.4 Selection and Training Training of Personnel Personnel The selection selection and training of operations operations personnel meets or exceeds exceeds the requirements of of ANSI!ANS-15.4-1988 Sections 4-6 1.'. The operator training program includes sufficient ANSI/ANS-15.4-1988 radiation safety training to meet the requirements requirements of 10 CFR Part 19 and the campus Radiation Radiation Safety Regulations. The operator training program program is a two step process. First the candidate must take an elective elective four credit-hour course with a formal training manual, homework, and practical exercises (On the Job Training) included. It includes the equivalent of 0.5 weeks weeks ofof reactor fundamentals, 1.25 weeks weeks of systems coverage, 0.5 weeks of systems observation, observation, and 0.8 weeks weeks of control room operations administered over the period of one semester. The course is completely completely described on the UW Nuclear Reactor web page. UWNR UWNR Safety Safety Analysis l U NRSaey Analysis Report ev Rev. 22 Report Rev. 1-4Spt 12-4 12-4 nayisReot 20 Sept. 2008 2008
*
- After completing completing the course, successful candidates are selected to participate participate in the candidacy candidacy program. Under Under this program, candidates perform approximately approximately 12 additional additional weeks of control room operations and other non-licensed non-licensed duties, including preventative preventative maintenance maintenance and health physics surveys.
The operator operator proficiency proficiency maintenance maintenance program (re-qualification (re-qualification program) fully meets the requirements of 10 CFR Part 55 and is formalized as a facility procedure, procedure, UWNR 004 2, 2, which received NRC approval upon initial implementation implementation and is reviewed annually annually by the facility organization along with other facility procedures. The program includes operating organization includes written, oral, and performance performance testing as well as emergency procedure emergency procedure drills and classes on changes in experiments, facility equipment, and procedures. experiments, 12.1.5 Radiation Safety Radiation safety aspects aspects of facility operation operation are routinely performed by members of the reactor reactor operating staff, including routine radiation and contamination contamination surveys and sampling sampling of water and air samples. The campus campus radiation safety organization (see chapter chapter 11), 11), established established to oversee all activities involving ionizing radiation radiation on campus, is part of the University Department of ofthe of Environment, Health and Safety, and thus is an independent organization independent organization which reports to the central central campus administration. The radiation safety organization organization has the authority to interdict or terminate terminate radiation safety related activities conducted under the reactor license license..
- UWNR Safety Analysis Report Rev. 2 12-5 12-5 2008 Sept. 2008
CHANCELLOR.: BOARD OF REGENTS CHANCELLOR- MADISON _i;,i'J[1TI£:~j'/f:_Ir ."~2:Q;?~* REGENTS MADISON CAMPUS CAMPUS
- UNIVERSITY OF WISCONSIN UNIVERSITY RADIATION RADIATION SAFETY COMMITTEE COMMITTEE m
.. :: ..
if!;JJ,;l?j,/ .. ",,' I University University Safety Department Department Radiation Radiation Safety OfficeOffice CHAIRMAN ENGINEERING PHYSICS DEPARTMENT ENGINEERING DEPARTMENT (ANSI/ANS-15.4 Level (ANSI/ANS-15.4 1) Level 1) REACTOR SAFETY REACTOR COMMITTEE SAFETY COMMITTEE ,--:-
- REACTOR DIRECTOR REACTOR DIRECTOR (ANSI/ANS-15.4 Level 2)
(ANSI/ANS-15.4 REACTOR REACTOR SUPERVISOR (SRO) SUPERVISOR (SRO) (ANSI/ANS-15.1 Level (ANSI/ANS-15.1 3) Level 3) ALTERNATE AL SUPERVISORS (SRO) TERNATE SUPERVISORS (SRO) (ANSI/ANS-15.1 (ANSI/ANS-1 5.1 Level 3) REACTOR OPERATORS REACTOR (RO) OPERATORS (RO) (ANSI/ANS-1 5.1 Level 4) (ANSI/ANS-15.1 Figure 12-1 UWNR 12-1 Organization Chart Organization Analysis Report Rev. 2 UWNR Safety AnalysisJReport 12-6 12-6 Sept. 2008 2008
*
- 12.2 Review and Audit Activities Qualifications 12.2.1 Composition and Qualifications The Reactor Reactor Safety Committee Committee is appointed by the Reactor Reactor Director. Minimum committee committee size is six members, one of whom is a health physicist from the University Radiation Safety office.
Other members are faculty and staffstaff of the university selected selected based on expertise expertise to assure that the following disciplines disciplines are represented: represented: 1.
- 1. Reactor Physics - Nuclear Engineering Engineering
- 2. Mechanical Engineering - Heat transfer Mechanical transfer and fluid mechanics 3.
- 3. Metallurgy/Materials Metallurgy/Materials
- 4. Instruments and Control Systems 5.
- 5. Chemistry Chemistry and Radio-chemistry Radio-chemistry
- 6. Radiation Safety Reactor operations operations staff is not precluded precluded from membership membership on the committee committee as long as such
- members do not reach members reach a majority of a quorum for voting. The health physics personnel perform the monthly audits audits and inspections are invited to the meetings, members of the committee.
12.2.2 Charter 12.2.2 Charter and Rules personnel who meetings, but are not necessarily The Reactor Safety Committee operates with a written charter charter which which specifies the manner manner in which business business is conducted. The charter includes rules on meeting frequency (at least annually), voting rules, agenda, quorums, use of subcommittees, subcommittees, minutes, and methods methods and content of of submissions to the committee. Provisions for use of telephone polls or subcommittees subcommittees for approval approval of items not requiring a formal meeting are also a part of the charter. 12.2.3 Review Function The reactor director or designee reviews all written operating procedures procedures atleast at least annually. Results of this review, along with suggested procedure procedure revisions, are submitted submitted to the Reactor Safety Safety Committee Committee for approval, or re-affirmation re-affirmation if no changes changes are deemed necessary. The review responsibilities responsibilities of the Reactor Safety Safety Committee shall include, but are not limited to, the following:
- UWNR Safety Analysis Report Rev. 2 12-7 12-7 2008 Sept. 2008 Sept;
1. 2. Review and approval of new experiments Review Review Review and approval experiments utilizing the reactor facilities. approval of all proposed changes and technical specifications. specifications. changes to the facility, procedures, procedures, license,
- 3.
- 3. Determination Determination of whether a proposed proposed change, test or experiment experiment would constitute an unreviewed unreviewed safety question or a change in Technical Specifications.
Specifications.
- 4. Review Review of abnormal performance performance of plant equipment and operating operating anomalies having safety significance.
5.
- 5. Review of unusual or reportable occurrences occurrences and incidents which are reportable reportable under 10 CFR Part 20 and 10 CFR Part 50.
- 6. Review of audit reports.
- 7. Review of violations of technical technical specifications, license, license, or procedures procedures and orders having safety safety significance.
significance. 12.2.4 Audit Function 12.2.4 Function Radiation Safety Office represents A Health Physicist from the University Radiation represents the University
- Radiation Radiation Safety Safety Committee and conducts an inspection of the facility at least monthly to assure compliance compliance with the regulations of 10 CFR Part 20. The services and inspection inspection function of the Reactor Safety Committee, with the scope of the Radiation Safety Office are also used by the Reactor Radiation extended to cover audit extended cover license, technical specification, and procedure procedure adherence.
adherence. 12.3 12.3 Procedures Procedures Written operating procedures Written operating procedures are used to assure the safety of operation operation of the reactor. Procedure use does not preclude the use of independent independent judgement judgement and action should should the situation require such. Operating procedures procedures are in effect for the following items: 1.
- 1. Testing and calibration calibration of reactor operating instrumentation and controls, control operating instrumentation control rod drives, area radiation radiation monitors, and air particulate particulate monitors.
- 2. Reactor startup, operation, and shutdown.
3.
- 3. Emergency Emergency and abnormal conditions, including provisions for evacuation, reentry, recovery, and medical support.
- 4. Fuel element and experiment experiment loading or unloading.
UWNR Safety UWNR Safety Analysis Analysis Report Report Rev. 2 12-8 12-8 2008 Sept. 2008 *
- 55.. Control rod removal or r~placement.
replacement. 6.
- 6. Routine maintenance Routine maintenance of the control rod drivesdrives and and reactor safety safety and interlock interlock systems or other routine routine maintenance maintenance that could have have an effect effect on reactor safety.
reactor safety. 7.
- 7. Actions Actions to be taken taken to correct correct specific specific and foreseen foreseen potential potential malfunctions malfunctions of systems systems or components, components, including including responses responses to alarms and abnormal abnormal reactivity reactivity changes.
changes. 8.
- 8. disturbances on or near Civil disturbances near the facility site.
Substantive changes Substantive changes to the above procedures procedures may be made only with the approval of the Reactor Reactor Safety Committee. Safety Committee. Temporary changes to the procedures procedures that do not change change their original original intent may be made approval of two SROs. All made with the approval temporary changes All such temporary documented and changes are documented and subsequently reviewed by subsequently reviewed Safety Committee. by the Reactor Safety Committee. 12.4 12.4 Required Actions Required Actions In the event event a safety safety limit is exceeded: exceeded: 1.
- 1. The reactor reactor shall be shut down and reactor reactor operation operation shall not be resumed until authorized by NRC..
by the NRC
- 2. immediate report of the occurrence An immediate Safety Committee, 6.7 of the technical Section 6.7 Section occurrence shall be made to the Chairman, Committee, and reports shall be made to the NRC specifications.
technical specifications. NRC in accordance Reactor Chairman, Reactor accordance with 3.
- 3. A report shall A prepared which shall be prepared which shall shall include an analysis analysis of the causes causes and extent of possible resultant resultant damage, efficacy corrective action, efficacy of corrective action, and recommendations recommendations measures to prevent or reduce for measures reduce the probability probability of recurrence.
recurrence. This report shall submitted to the Reactor Safety be submitted Safety Committee Committee (RSC) for review and then submitted submitted to the NRC when when authorization authorization is sought to resume operation operation of the reactor. A reportable occurrence A defined as any of the following occurrence is defined following that that occur during reactor reactor operation: 1.
- 1. Operation with any safety Operation safety system setting less conservative system setting conservative than specified specified in the technical technical specifications.
specifications.
- 2. Operation in violation Operation Limiting Condition violation of a Limiting Operation listed in the Condition for Operation Technical Specifications Technical Specifications..
- UWNR UWNR SafetySafety Analysis Analysis Report Rev. 2 12-9 129Sp.08 12-9 2008 Sept. 2008
3. 4. reactor or experiment safety system component Operation with a required reactor inoperative or failed condition which could render inoperative performing its intended safety function. unanticipated or uncontrolled Any unanticipated component in an render the system incapable of uncontrolled change in reactivity greater than 0.7% of
~KlK, 0.7% AK/K,
- excluding reactor trips from a known cause.
excluding
- 5. An observed observed inadequacy in the implementation of either administrative or procedural controls, such that the inadequacy procedural existence or inadequacy could have caused the existence development of a condition which could result in operation of the reactor outside the specified safety limits.
, 6. degradation in reactor fuel or cladding Abnormal and significant degradation cladding which could result in exceeding prescribed radiation exposure limits of personnel or exceeding prescribed environment, or both.
occurrence as defined in the Technical Specifications, In the event of an reportable occurrence Specifications, the following actions shall be taken: 1.
- 1. The reactor shall be shut down.
alternate shall be notified and corrective Director or designated alternate
*
- 2. The Reactor Director action taken with respect to the operations involved.
- 3. The Director or designated alternate shall notify the Chairman of the ReactorReactor Safety Committee.
- 4. A report shall be made to the Reactor Safety Committee which shall include an analysis cause of the occurrence, analysis of the cause occurrence, efficacy corrective action, and efficacy of corrective recommendations for measures to prevent recommendations prevent or reduce the probability of recurrence.
5.
- 5. A report shall be made to the NRC.
12.5 12.5 Reports Reports will be made to NRC in accordance with the following: 1.
- 1. An annual report covering the activities of the reactor facility during the previous report covering previous calendar calendar year shall be submitted (in writing to U.S. Nuclear Regulatory Document Control Desk, Washington, DC 20555)
Commission, Attn: Document 20555) within six six months following the end of each calendar year, providing the following information: Analysis Report Rev. 2 UWNR Safety Analysis 12-10 12-10 2008 Sept. 2008
*
- a.
- a. A summary of (1)
A brief narrative summary (1) operating operating experience experience (including (including experiments performed), experiments performed), (2) changes changes in facility design, performance performance characteristics, and operating characteristics, operating procedures procedures related to reactor safetysafety and and occurring during the reporting occurring reporting period, and (3) (3) results of surveillance surveillance tests inspections. and inspections. b.
- b. Tabulation of the energy output (in megawatt days) of the reactor, hours Tabulation hours reactor reactor was critical, and the cumulative was critical, cumulative total energy output since initial criticality.
criticality. C.
- c. The number number of emergency shutdowns and inadvertent emergency shutdowns scrams, including inadvertent scrams, including reasons reasons therefor.
therefor. d.
- d. Discussion of the major maintenance maintenance operations performed during the operations performed period, including-the period, including-the effect, if any, on the safety operation of the safety of the operation
_*reactor reactor and the reasons reasons for any corrective maintenance required. corrective maintenance required.
- e. A brief description, A description, including summary of the safety including a summary evaluations of safety evaluations of changes in the facility or in the procedures changes procedures and of tests and experiments experiments carried carried pursuant Section 50.59 pursuant to Section 50.59 of 10 10 CFR Part 50. 50.
ff.. A summary of the nature and amount of radioactive effluents effluents released or
- A summary discharged discharged to the environs environs beyond beyond the effective effective control control, of the the licensee licensee as measured measured at or prior to the point of such release or discharge.
i.
- 1. Liquid effluents (summarized Liquid effluents (summarized on a monthly basis)
(1) (1) Liquid radioactivity radioactivity discharged discharged during the reporting period.-period. Tabulated Tabulated as follows: (a) Total estimated radioactivity Total estimated radioactivity released released (in curies). (b) (b) The isotopic isotopic composition greater than 1 x 10-composition if greater 10'-7 microcuries/cc microcuries/cc for fission and activation activation products. products. (c) Total radioactivity released by curies), released radioactivity (in curies), by nuclide, nuclide, during the reporting reporting period period based on representative representative isotopic isotopic analysis. (d) (d) Average concentration at point of release Average concentration release (in (in microcuries/c) microcuries/cc) during the reporting period period and and the fraction fraction of the applicable applicable limit in 10 10 CFR 20.20 .
- UWNR UWNR Safety Analysis Report Report Rev. 2 1-1Sp.20 12-11 12-11 Sept. 2008 2008
ii. 11. (2) (2) Gaseous Total volume (in gallons) of effluent water (including diluent) during periods of release. Gaseous Waste (summarized (summarized on a monthly monthly basis)
*
(1) Radioactivity Radioactivity discharged during the reporting period (in curies) for: (a) Gases. (b) Particulates Particulates with half lives greater than eight days. The estimated estimated activity (in curies) discharged discharged during the reporting reporting period, by nuclide, for all gases and particulates particulates based based on representative representative isotopic analysis and the fraction ofof applicable 10 CFR 20 limits for these values. the applicable iii. 111. Solid Waste (1) The total amount of solid waste packaged packaged (in cubic feet).
*
(2) (2) The total activity involved (in curies). (3) The dates of shipment and disposition (if shipped off site).
- g. A summary summary of radiation exposures received by facility personnel and visitors, including dates and time of significant significant exposures exposures and a summary of the results of radiation and contamination contamination surveys surveys performed performed within the facility.
- h. A description of any environmental environmental surveys performed performed outside the facility.
- 2. A report within 60 days after completion of startup testing of the reactor (in writing to the U.S. Nuclear Regulatory Regulatory Commission, Attn: Document Control Desk, Washington, D.C. 20555 with a copy to the NRC compliance inspector compliance inspector assigned assigned to the facility) upon receipt of a new facility license or an amendment to the license authorizing authorizing an increase increase in reactor power level describing the measured measured values of the operating conditions or characteristics characteristics of the reactor reactor under the new conditions conditions including:
- a. An evaluation evaluation of facility performance performance to date in comparison with design predictions predictions and specifications.
specifications. UWNR UWNR Safety Analysis Report Rev. 2 12-12 12-12 2008 Sept. 2008
*
- b..
b reassessment of the safety analysis submitted with the license A reassessment measured operating application in light of measured application operating characteristics such characteristics when such measurements indicate that there may be substantial variance measurements prior variance from prior analysis.
- 3. telephone or A report of any of the following not later than the following day by telephone conveyance to the NRC Headquarters similar conveyance Operation Center, and followed Headquarters Operation followed by a written report describing the circumstances circumstances of the event and sent within 14 days to U.S. Nuclear Regulatory commission, Attn: Document Control Desk, inspector assigned to the 20555, with a copy to the NRC inspector Washington, D.C. 20555, facility:
- a. Any accidental release of radioactivity above permissible limits in accidental release unrestricted areas whether or not the release resulted in property unrestricted property damage, personal injury, or exposure.
- b. violation of a safety limit.
Any violation
- c. occurrences.
Any reportable occurrences.
- 4. A written report within 30 days in writing to the U.S. Nuclear Regulatory Commission, Attn: Document Control Desk, Washington, D.C. 20555, of: of:
- a.
a. b. Department Chair level. Any significant Analysis Report. the Safety Analysis organization at Reactor Director or Permanent changes in facility organization Permanent or accident analysis as described in change in the transient or accident significant change 12.6 12.6 Records The following records are retained for a period of at least five years or for the life of the involved if less than five years. component involved 1.
- 1. Normal reactor facility operation (but not including supporting documents such as maintained for a period of at least one checklists, log sheets, etc. which shall be maintained checklists, year)
- 2. Principal maintenance activities Principal maintenance activities
- 3. Reportable occurrences occurrences
- 4. Surveillance Surveillance activities required by the Technical activities required Specifications Technical Specifications
- UWNR Safety Analysis Report Rev. 2 12-13 12-13 .. Sept. 2008 2008
5. 5. 6. Reactor facility radiation and contamination regulations regulations contamination surveys where required by applicable Experiments performed with the reactor reactor applicable
*
- 7. Fuel inventories, inventories, receipts, receipts, and shipments shipments
- 8. Approval of changes procedures changes in operating procedures
- 9. Records of meeting meeting and audit reports of the review and audit group.
Operator Operator qualification and re-qualification records will be retained for at least one cycle cycle of the re-qualification program. The following records records will be retained for the lifetime of the reactor facility. (Note: Retention of of annual annual reports which which contain the information in items 1.
- 1. and 2. are considered considered as suitable records records for those items.)
1.
- 1. Gaseous and liquid radioactive radioactive effluents released to the environs, environs
- 2. Offsite environmental environmental monitoring surveys required by technical specifications specifications 12.7 3.
4. Emergency Radiation Radiation exposures for all personnel monitored Updated, corrected, Emergency Planning corrected, and as-built drawings of the facility Planning
- The Emergency Plan for the University University of Wisconsin Nuclear Reactor was prepared prepared to meet the requirements requirements of ANSI/ANS 15.16-1978 3 as amplified by Nureg-0849 4.'. This plan was submitted ANSI/ANS 15.16-1978 3 submitted to NRC for review in May 21,1980, 21,1980, with subsequent revisions in October October 25, 1982, and May 17, 1984. By letter dated July 25,1984 25, 1984 NRC indicated that the plan met the requirements referenced referenced above. The plan was again modified and submitted to NRC on May 16, 1990, 1990, with supporting submaitted on August 12, 1990. NRC notification that the revision was acceptable
. information subinitted acceptable was received in a letter dated April 26, 1991. 1991. The plan was again modified (Revision 4) to reflect reflect the changes changes in section section number nomenclature of 10 CFR Part 20 and submitted to NRC on number and nomenclature February 17, February] 7, 1994 and April 22, 1994. This version is the current version in use at the facility. The Emergency Emergency Plan indicates indicates response capabilities for emergency emergency conditions arising in connection connection with operation operation of the reactor. It includes identification identification of various precursor conditions (loss of electrical power, fires, reactor reactor pool leaks, riots, etc) and the consequences for various independent independent or simultaneous precursor. The plan includes the event classification classification system. The 0 dose to which people could be exposed under various conditions is indicated, as are the actions UWNR Safety Analysis Analysis Report Rev. 2 12-14 12-i4 2008 Sept. 2008 *
- that can be Delegation be taken to minimize implementing procedures implementing Primary responsibility minimize the consequences responsibility for emergency Delegation of responsibility Emergency Plan and implementing The Emergency consequences of emergency planning the emergency.
ofthe emergency. Detailed procedures have been developed and are referenced planning and response is given to the Reactor responsibility and authority authority in the absence implementing procedures absence of the Reactor procedures are reviewed emergency Detailed emergency referenced in the plan. Reactor Director. annually to assure reviewed annually Director. Reactor Director is specified. specified. assure that any any required required changes incorporated into the plan. changes are incorporated 12.8 12.8 Security Planning Planning The facility security plan, UWNR facility physical security UWvNR 003,003, was initially initially submitted to NRC on October 18, October 18, 1988. The security plan was revised 1988. submitted again on June 17,1991 revised and submitted 17,1991 and was found to meet requirements, including the applicable requirements, including the format of Regulatory Regulatory Guide 5.59 '.5. The plan Guide 5.59 plan will will require revision as a result of this Safety Analysis Report, since some figures from the previous Safety Analysis previous Safety Analysis included by Analysis Report are included reference. These by reference. changes will be made during These changes during the usual annual reviews of the plan once this SAR becomes referenced in the reactor becomes the document referenced reactor operating license. operating license. The security security plan indicates the measures provided provided to protect protect special nuclear nuclear material, material, including including details of the protective equipment and police agencies, and is thus withheld protective equipment from public disclosure. Reactor Director is responsible disclosure. The Reactor responsible for administering administering the security security program program and
- assuring that it is updated updated as required.
required . 12.9 12.9 Quality Assurance construction permit is sought Since no construction sought in the application application for renewal renewal of the license for the
.University of Wisconsin description of a quality Wisconsin Nuclear Reactor, no description quality assurance assurance program for the design and construction components of the facility is included.
construction of the structures, systems, and components included. section describes This section describes the Quality Assurance program Assurance program that is in place to govern place govern safe operation operation modification of the facility. This program and modification program meets the applicable applicable requirements requirements of Regulatory Regulatory Guide 2.5 6and ANSI/ANS-6 15.8-1995' ANSIIANS-15.8-1995 7 Director has responsibility The Reactor Director responsibility for the quality assurance assurance activities, and thus has the authority to identify problems, to initiate corrective corrective actions, and to insureinsure that corrective corrective actions are performed. QA oversight by exercises QA performed. He exercises operating and maintenance by assuring that operating maintenance procedures include specific procedures specific requirements to assure modification, maintenance, assure that modification, maintenance, and calibration of safety-related calibration performed in a manner that maintains the quality and safety-related systems are performed reviews use written requirements experiment reviews reliability of equipment. Further, experiment reliability requirements to assure that installation installation and operation of the experiment experiment does not degrade the performance performance of safety equipment. Modification of safety-related equipment. Modification equipment is planned safety-related equipment reviewed using formal planned and reviewed written procedures that assure that equipment written checklist-type procedures equipment continues continues to meet the original specifications. Most ofthe specifications. of the reactor reactor equipment equipment in use in the facility does not have formal QA QA documentation because it was built before documentation because before the QAQA requirements requirements were in effect. This equipment equipment is
- UWNR Safety Analysis Report UWNR Report Rev. 2 12-15 1-5Sp.20 12-15 2008 Sept. 2008
covered covered under the provisions provisions of section 4 of ANSI/ANS-15.8. ANSI/ANS-15.8. Several instruments instruments are
- replacements electronics originally provided by the reactor replacements for the vacuum-tube electronics reactor manufacturer, General General Electric Company. This replacement replacement equipment was designed, built, and tested to meet the original specifications specifications stated in the equipment manualsmanuals provided with the General Electric equipment. After-maintenance checks, After-maintenance checks, alignment, and calibration calibration of the replacement equipment replacement equipment still assures the equipment equipment meets the original equipment equipment specifications.
Procedures include schedules of equipment Procedures maintenance and calibration, and provide equipment maintenance provide records records that such functions have been completed. Calibration procedures procedures include include requirements requirements that critical equipment and instruments instruments used in the calibrations calibrations are themselves themselves currently calibrated calibrated (when appropriate).). appropriate 12.10 Operator Operator Training Training and Requalification Requalification Operator Operator training and Requalification Requalification programs are briefly described in section section 12.1.4. 12.1.4. The Requalification Requalification plan at UWNR is published as a standard procedure (UWNR (UWNR 004, "University of "University of Wisconsin Wisconsin Nuclear Reactor Operator Proficiency Maintenance Maintenance Program") Program") which was submitted to NRC on October 24, 1973 and revised revised on February 7, 1974. By letter February letter dated March 1974 March 29, 1974 we were notified by NRC that the program meets the requirements requirements of Section Section 50.54(i-1) 50.54(i-l) of 10 10 CFR Part 50 and Appendix A of 10 CFR Part 55. Since the program program is a numbered numbered procedure procedure it is reviewed reviewed by management management on an annual basis. 12.11 Startup Plan The facility has been in routine operation for many years, so a startup Safety Analysis Report for license renewal. startup plan is not included in this
- Environmental Reports 12.12 Environmental On January 23,23, 1974 the AEC staff concluded concluded in a memorandum Skovholt and memorandum addressed to D. Skovholt and signed by D.D. R. Miller, "that there will be no significant environmental environmental impact associated with the licensing licensing of research reactors reactors or critical facilities designed to operate operate at power levels of 2 Mwt or lower and that no environmental environmental impact statements statements are required to be written for the construction permits or operating licenses for such facilities."
issuance of construction Since this Safety Analysis Analysis Report is written in support support of extending the license expiration date for an additional additional 20 years, no changes changes in land and water use are contemplated. Emissions of of radioactive materials materials or other effluents will not change as a result of extending extending the license term. 12.13 References 12.13 References 1.
- 1. ANSI/ANS-15.4-1988, Selection Standard ANSIIANS-15.4-1988, Selection and Training Training of Personnel For Research Reactors, American American Nuclear Society, June 9, 1988 1988 ANSI Approval UWNR Safety Analysis Report Rev. 2 UWNR 12-16 12-16 2008 Sept. 2008 *
- 2.
3. UWNR 004, Operator Proficiency UWNR ANSIIANS ANSI/ANS 15.16-1978, Proficiency Maintenance Maintenance Program Reactors", ANS, LaGrange 15.16-1978, "Emergency; Planning for Research Reactors", Park, Illinois, 1978 LaGrange
- 4. NUREG-0849, "Standard Review Plan for the Review NUREG-0849, Review and Evaluation Evaluation of Emergency Emergency Plans for Research Research and Test Reactors", USNRC, October 1983 Reactors", USNRC, 1983 5.
- 5. Regulatory Guide Regulatory Guide 5.59, Revision Revision 1, "Standard Format and Content for A Licensee Licensee Physical Security Plan for the Protection Special Nuclear Protection of Special Nuclear Material Material of Moderate Moderate or Low Low Strategic Significance, Significance, US Nuclear Regulatory Regulatory Commission, February February 1983 1983
- 6. Regulatory Regulatory Guide 2.5, Revision O-R,0-R, :Quality
- Quality Assurance Assurance Program Program Requirements Requirements for for Research Research Reactors, October 1977 1977
- 7. ANSI-15.8-1995, "quality ANSI-15.8-1995, "quality assurance program requirements for research reactors", ANS, La Grange Grange Park, Illinois, 1995 1995
- UWNR Safety Analysis UWNR Analysis Report Rev. 2 12-17 12-17 Sept. 2008 2008
- intentionally left blank.
This page is intentionally
- UWNR Safety Analysis Report Rev. 2 12- 18 12-18 2008 Sept. 2008
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- 13 13 ACCIDENT ANALYSIS ACCIDENT ANALYSIS NUREG -1537-1537 1' divides accident accident analysis into initiating initiating and consequences sections. In this report the two sections are combined; combined; that is, the analyses analyses of the consequences consequences of accidents are grouped with the accident-initiating accident-initiating events and scenarios scenarios in order to reduce reduce duplication duplication which would otherwise occur. In addition, a better appreciation of the likelihood better appreciation likelihood and consequences consequences of of accidents is afforded.
afforded. The sections are numbered to correspond correspond to the numbering system of of NUREG-1537. NUREG-1537. 13.1 13.1 Initiating Events and Determination Accident Analysis Initiating Determination of Consequences Note: NUREG-1537 NUREG-1537 specifies separate sections for Initiating Events and Determination of of Consequences. Both of these sections have Consequences. have been combined for clarity. NUREG/CR-2387 2 reports an independent study of accidents in TRIGA-type reactors and NUREG/CR-2387 concludes "The only potential concludes "The potential for offsite offsite exposure appears appears to be from a fuel-handling accident that, based on highly conservative equivalents of *< conservative assumptions, would result in dose equivalents .s;; 1 mrem to the total body from noble gases and .s;;__1.2 rem to the thyroid from radioiodines." Notwithstanding radioiodines." Notwithstanding conclusion, the following that conclusion, following sections sections reiterate reiterate the analysis analysis done for a license amendment amendment allowing the University of Wisconsin Nuclear Reactor Reactor to operate with FLIP or mixed Standard-FLIP cores, using values specific to the UWNR reactor location and characteristics. characteristics.
- 13.1.1 Maximum Hypothetical l3.1.1 Hypothetical Accident The maximum maximum hypothetical accident for UWNR UWNR is postulated as damage to a fuel element resulting in failure of the fuel cladding. It is postulated postulated that this damage occurs after a very long time of operation 125% of full power (the power operation at 125% power level limiting safety system setting) and that it occurs in the fuel element with the highest power density possible in permitted permitted UWNR mixed mixed core fuel loadings. In a compact 9-Bundle FLIP core the highest power power density is 18.1 kW kWatat 1000 kW; the corresponding corresponding number number for the 15-bundle FLIP core is 17.2 kW, while the value for currently-used all FLIP core is only 15.2 kW. Continuous the currently-used Continuous operation at the power level scram setpoint is highly unlikely, but for this hypothetical hypothetical computation the power level in the maximally exposed fuel element of the 9-bundle maximally 9-bundle core is assumed to be 23 kW (1.25 (1.25 times 18.l18.1 kW rounded to the next highest number).
The likelihood of a major major fuel element cladding failure is considered considered small. The elements must meet rigid quality control standards; standards; pool water quality is carefully carefully controlled; and much care is taken in handling fuel. Such clad failures are, however, possible and the remainder remainder of this section is concerned with the consequences consequences of such a failure. The release release of radioactivity by corrosion and leaching by the pool water has been measured measured at Gulf General Atomic. About 100 micrograms micrograms of U-ZrHU-ZrH per square centimeter exposed fuel centimeter of surface surface per day is released for shutdown shutdown conditions. This releaserelease is easily controlled by isolating
- UWNR Safety Analysis Analysis Report Report Rev. 2 13-1 13-1 2008 Sept. 2008
the leaking element element in a container container provided for that purpose. The gaseous and highly volatile
- fission products that have collected in the space between fuel and cladding would be the activity contributing contributing to personnel hazards.
13.1.1.1 Fission Product Inventory in Fuel Element The quantity quantity of these volatile and gaseous fission products products was determined by the use of Perkins 33 and King data. Column B of Table 13.1 indicates the fission product activities in the fuel element exposed to the maximum power density. 13.1.1.2 13.1.1.2 Fission Product Release Release Fraction The release of fission products from U-ZRH fuel elements has been extensively studied by Gulf Gulf General General Atomic and others. The results of this work indicate that the release of fission product product gases into the gap between fuel and cladding is given by the following relationship: FR=1.5E-5 FR=1.5E-5 + 3.6E3exp(-1.34E4/T) 3.6E3exp(-1.34E4/T) where T is the maximum fuel temperature ('K) in the element during normal operation. temperature (DK) The maximum fuel temperature in a fuel element element operated in the steady-state steady-state mode at 23 KW will be less than 440 DC.'C. Calculations of release fraction however, are based on 600 'C DC in order
- to assure a conservative conservative result.
The release fraction corresponding corresponding to 600 DC 'C is 7.9 E-4. Applying this fraction to the total inventory of the fuel element as given in column Table, t3.1 column B of Table' 03.1 gives the released activity as , shown in column column C of the table. For the purpose of further calculations, it is assumed that all gaseous fission products products are released to the room air whether the pool is filled with water or not. For soluble volatiles, calculations calculations assume all activity is absorbed absorbed in pool water for calculations calculations of pool water activity (column D). For calculations of air activity, the assumption is made that 10% 10% of the volatiles escape with the pool filled with water (columns E and F) and 100% 100% escape with the pool empty. 13.1.1.3 13.1.1.3 Activity in Pool Water Activity Water If 100% 100% of the soluble fission products products are absorbed in the pool water, the resulting activity 'level level will be 0.075 [tCi/ml. IlCi/ml. Within 24 hours the level would be reduced by radioactive decay to about 0.012 IlCi/ml. p.Ci/ml. After 24 hours the activity decay rate would be chiefly determined by the 1-131 1-131 half life (8.05 days). The halflife The demineralizer demineralizer will remove most of this activity, giving a radiation dose rate of about 88 mrem/hr mrem/hr one meter at meter after the activity activity is deposited in the resins. The resins can can be dumped to an underground underground storage pit or the underground underground liquid waste holdup tank where the activity will decay without hazard hazard to personnel. UWNR UWNR Safety Analysis Report Rev. 2 13-2 13-2 Sept. 2008 2008 *
- 13.1.1.4 13.1.1.4 Fission Product Product Release to Air within the Reactor Laboratory It is estimated that it would take an individual five minutes to evacuate confinement confinement and an additional five minutes to evacuate evacuate the building. However However for conservatism, conservatism, it is assumed that the individual individual is exposed at the highest concentration, namely that found in confinement, for the full ten minutes. Calculations Calculations were performed in Appendix A to determine (1) the dose rate due to gamma emitters emitters uniformly dispersed throughout the volume of the reactor lab, (2) (2) the dose to the lungs from beta emitters for an individual remainingremaining in the laboratory laboratory for ten minutes; and (3) the dose to the thyroid of an individual remaining remaining in the room ten minutes. For the latter latter calculations, calculations, it is assumed that 10% of 10% ofthe the iodine radioisotopes escape from the pool water. In In addition to these calculations, a computation of the number ofDAC of DAC hours indicates that a person person present in the room for 10 minutes minutes after the release would receive less than the annual limit on intake for occupational exposure.
(1) (1) Whole body exposure exposure due to gamma gamma emitters The amount of insoluble volatiles volatiles released to the room would be 5.89 Ci. If this activity is distributed uniformly in the laboratory concentration would be 2.95E-3 laboratory volume, the resulting concentration I.Ci/cm 3 3 IlCi/cm *. The resulting maximum dose rate is calculated calculated to be 60 mrem/hr. An individual laboratory for 10 minutes after a release would receive a whole body dose of 10 remaining in the laboratory 10 mrem.. mrem
- (2)
(2) Dose to the lungs The lung is the critical organ when considering considering the effects of inhaling the insoluble a ruptured fuel element. The beta emitting nuclides gamma become nuclides become more insoluble volatiles from important than those emitting gamma rays since all the decay energy is absorbed in lung tissue. The calculation outlined in the appendix appendix indicates the lung exposure exposure for an individual remaining remaining in the laboratory for 10 minutes after a clad rupture rupture to be 1.0 rad. (3) (3) Thyroid Thyroid dose The thyroid dose to a person in the reactor calculated assuming that he remained in the reactor room was calculated laboratory for 10 minutes after the fission product release. If the pool water is not lost and 10% 10% of the halogens halogens released escape into the atmosphere, atmosphere, the concentrations concentrations of the various iodine isotopes would be as presented in Table 13.1. 13.1. In a ten minute period the lungs would be exposed exposed to the iodine isotope activities activities shown in Table 13.2. As before, before, it was assumed that the "standard man" 3 "standard man" breathes breathes 1.25 1.25 mm 3/active-hour
/active-hour andand his lungs hold 3 liters of air. A conservative his lungs conservative calculation results in a dose to the thyroid of 18.9 rads. Although calculation Although all doses were calculated based based on an individual individual remaining remaining in the laboratory laboratory for ten minutes, emergency emergency procedures procedures require immediate immediate evacuation evacuation after scramming scramming the reactor, and re-entry to the area is made using a purifying respirator. Actual powered air purifYing Actual doses in the event of the accident would be a factor of of 10 less than calculated, calculated, considering evacuation times considering reasonable evacuation times..
- UWNR UWNR Safety Analysis Report Rev. 2 13-3 13-3 2008 Sept. 2008
- TABLE 13.1 FISSION PRODUCT RELEASE FROM PRODUCT RELEASE FROM CLAD RUPTURE
*
- A B C D E F G H H I J ISOTOPE SATURATED SATURATED RELEASED AMOUNT AMOUNT IN IN AMOUNT 1N IN LABORATORY LABORATORY Part 20 CONCENTRATION CONCENTRATION Part 20 TABLE 2 RATIO INVENTORY WATER INVENTORY INVENTORY INVENTORY (Ci)
WATER(Ci) AIR AIR (Ci) CONCENTRATION CONCENTRATION DISCHARGED ([.tCi/ml) TABLE 1 DISCHARGED (!lCi/ml) MAX.EFFLUENT MAX. EFFLUENT H/I COL. HlI (Ci) (Ci) (ItCi/ml) (!lCi/ml) DAC CONCENTRATION CONCENTRATION (-tC i/ml) (!lCi/ml) (4tCi/ml) (!lCi/ml) Br 82 82 30 0.024 0.024 0.024 0.002 1.2E-06 2E-06 1.46E-10 1.46E-10 6E-09 6E-09 0.0244 83 83 105 0.083 0.083 0.083 0.008 4.2E-06 4.2E-06 3E-05 4.95E-10 4.95E-10 9E-08 0.0055 84 84 194 194 0.153 0.154 0.154 0.015 0.015 7.7E-06 2E-05 9.23E-10 9.23E-10 8E-08 8E-08 0.0115 0.0115
'85
'85 253 0.200 0.200 0.020 1.OE-05 1.0E-05 1E-07 1E-07 1.20E-09 1.20E-09 IE-09 1E-09 1.2038 1.2038
'87
'87 600 0.473 0.475 0.047 2.4E-05 1E-07 1E-07 2.85E-09 1E-09 1E-09 2.8463 2.8463 TOTAL TOTALBr Br 0.933 0.093 I '130m
'130m 200 0.158 0.158 0.158 0.016 0.016 7.9E-06 1E-07 1E-07 9.45E-10 9.45E-10 1E-09 1E-09 0.9450 0.9450 131 131 563 0.446 0.446 0.446 0.045 0.045 2.2E-05 2E-08 2.68E-09 2.68E-09 2E-10 2E-10 13.33875 13.3875 132 132 855 0.677 0.677 0.677 0.068 0.068 3.4E-05 3E-06 3E-06 4.06E-09 4.06E-09 2E-08 0.2031 0.2031 133 133 1282 1282 1.015 1.015 1.015 1.015 0.102 0.102. 5.1E-05 1E-07 1E-07 6.09E-09 IE-09 1E-09 6.0863 134 134 1554 1554 1.230 1.230 1.230 1.230 0.123 6.2E-05 2E-05 7.38E-09 6E-08 0.1230 0.1230 135 135 1185 1185 0.938 0.938 0.938 0.094 0.094 4.7E-05 7E-07 7E-07 5.63E-09 6E-09 0.9375
'136
'136 602 0.477 0.477 0.477 0.048 0.048 2.4E-05 1E-07 1E-07 2.86E-09 2.86E-09 I E-09 1E-09 2.8575 TOTAL I 4.941 0.494 0.494 Kr 83m 83m 105 105 0.084 0.084 0.084 4.2E-05 IE-02 1E-02 4.95E-10 4.95E-10 5E-05 5E-05 9.9E-06 85m 85m 253 0.200 0.200 0.200 1.OE-04 1.0E-04 2E-05 1.20E-08 1.20E-08 1E-07 0.1204 85 51 51 0.040 0.040 0.040 2.OE-05 2.0E-05 1E-04 1E-04 2.40E-09 2.40E-09 7E-07 0.0034 0.0034 87 486 0.386 0.385 1.9E-04 1.9E-04 5E-06 2.3 1E-08 2.31E-08 2E-08 2E-08 1.1531 1.1531 88 88 699 0.556 0.555 0.555 2.8E-04 2E-06 3.32E-08 332E-08 9E-09 3.6875
'89
'89 855 0.669 0.669 0.669 3.4E-04 3.4E-04 1E-07 1E-07 4.06E-08 4.06E-08 1E-09 1E-09 40.6125 TOTAL TOTALKr Kr 1.935 1.935 1.935 1.935 Xe 131m 5 0.004 0.004 0.004 2.OE-06 2.0E-06 4E-04 2.36E-10 2.36E-10 2E-06 0.0001 133m 31 0.025 0.025 1.6E-05 1.6E-05 1E-04 1E-04 1.50E-09 1.50E-09 6E-07 0.0025 133 133 1282 1.015 1.015 1.015 1.015 5. 1E-04 5.1E-04 1E-04 1E-04 6.09E-08 6.09E-08 5E-07 0.1217 0.1217 135m 350 350 0.277 0.277 0.277 1.4E-04 1.4E-04 9E-06 1.67E-08 1.67E-08 4E-08 0.4163 135 135 1243 1243 0.984 0.984 0.984 4.9E-04 IE-05 1E-05 5.91E-08 5.91E-08 7E-08 0.8438 0.8438
'137
'137 1185 1185 0.938 0.938 0.938 4.7E-04 4.7E-04 IE-07 lE-07 5.63E-08 1E-09 lE-09 56.2500 138 894 0.707 0.707 3.5E-04 4E-06 4.24E-08 4.24E-08 2E-08 2.1206 TOTAL Xe TOTALXe 3.950 3.950 3.950 3.950 indicates generic <2 hr half-life values used
,'indicates used UWNR Safety Analysis Report Rev. 2 UWNR 13-4 2008 Sept. 2008
* *
- Continuation of Table 13.1 13.1 SELECTED RELEASE RELEASE TOTALS Halogen Gamma Emitters 5.2 Ci Ci Halogen Halogen Beta Emitters Emitters 5.8 Ci Ci Total Halogens Ci 5.87 Ci Insoluble Insoluble Gamma Emitters Emitters 3.52 Ci Insoluble Insoluble Beta Emitters 5.50 Ci Total Insoluble Volatiles Volatiles 5.89 Ci UWNR Safety Analysis Report Report Rev. 0 13-5 13-5 April 2000 2000
1 13.1.1.5 13.1.1.5 Release of Fission Products to Unrestricted Unrestricted Areas Columns H, I, and J of Table 13.1 are concerned with the exposure of personnel outside the restricted area. Calculations were performed as indicated in Appendix A. The maximum restricted concentrations which might be expected in unrestricted unrestricted areas were calculated under the
- assumption assumption that venting took place in (he the time required required for the ventilation ventilation system to make one complete change in the laboratory (1569 (1569 seconds). Wind velocity was assumed to be the lowest average for any month (3.54 m/s).
The total dose to personnel personnel in the unrestricted unrestricted area is independent of whether the ventilation ventilation system is operating in normal mode (one exhaust fan running) or emergency venting mode (both system (both exhaust fans running); the concentration exhaust concentration would be considerably higher in the case of emergency emergency venting mode, but the period of exposure would be proportionally proportionally shorter. It is also emphasized emphasized that the total exposure exposure figure is a maximum to be expected expected at any point other than within the areas evacuated in the event of an accidental release. The total of the ratios of instantaneous instantaneous individual concentrations to 10 CFR Part 20 Appendix B individual concentrations Table 2 maximum air concentrations Table concentrations for discharge (sum of column J from Table 13.1) 13.1) is calculated to be 134.0, where the maximum calculated maximum air concentration concentration values are for unrestricted areas, 168 hours per week (24 hours per day, 7 days per week). When averaged averaged over a year's time of of ventilation (instead of just 1569 seconds), the resulting average ventilation average concentration concentration is 0.007 of the maximum maximum indicated by 10 CFR Part 20 for non-occupational non-occupational exposure exposure in unrestricted unrestricted areas. Even with the effluent discharge from normal operation (see Section 11 .1.1) the total Section 11.1.1) concentration to which personnel concentration personnel might be exposed is below the 10 CFR Part 20 limits. A more conservative conservative calculation calculation which assumes zero stack height (see Appendix A) was performed. This analysis analysis is applicable to a situation in which the laboratory ventilation ventilation system system fails and the release release takes place through building leaks. For purposes of comparison, comparison, it was again assumed that the release occurred occurred in the time required for the ventilation ventilation system to make an air change in the laboratory ((1569 1569 seconds). The effect of this analysis is to multiply the values in in columns H and I by a factor of 2.6, giving a resulting average concentration of2.6, concentration (yearly average) of of 0.018 times 10 CFR Part 20 Appendix Appendix B Table 2 limits. Finally, an additional calculation calculation was performed performed assuming 100%100% release ofBrof Br and I and the more conservative calculation (zero stack height) of atmospheric conservative atmospheric dilution. The resulting resulting summation of of ratio of concentrations concentrations to release limit in this case would be 10391039.. Averaged Averaged over a year's time, the resulting concentration average) is 0.052 times the 10 CFR Part 20 Appendix concentration (yearly average) AppendixB B Table 2 limits, still within the permissible release concentration permissible release concentration when averaged over a period of one year. As indicated indicated in the table, releases are in all cases less than the 10 CFR Part 20 Appendix B Table Table 2 limits when averaged over one year. As a backup backup check to assure that these calculations calculations were conservative, conservative, cases equivalent equivalent to the m~ximum maximum hypothetical hypothetical acident were entered entered into the EPA UWNR Safety UWNR Safety Analysis Analysis Report Report Rev. 2 13-6 13-6 Sept. 2008 Sept.'2008
*
- COMPLY program 44 at level 4. Six of the radioisotopes COMPLY listing of radioisotopes included in the COMPLY COMPLY COMPL radioisotopes in Table 13.1 are not in the COMPLY (BR-85, Br-87, 1-130m, radioisotopes (BR-85, I-130m, 1-136, Kr-89, and Xe-137) and are COMPLY thus half-lives less than 10 minutes). Further, COMPLY results (these all have half-lives not Y does not permit zero height releases from a building, so the stack height was input as operable and the pool filled, COMPLY 1 meter. For the release with the ventilation system operable COMPLY averaged dose of 1.8 mrem, with 1.4 mrem due to Iodine. For the release indicated an annually averaged with pool water lost and the ventilation system inoperable (assumed stack height of 1 meter) the inoperable (assumed indicated an annual dose of 14.2 mrem and 13.7 mrem from Iodine.
COMPLY program indicated COMPLY 13.2 Table 13.2 Maximum Exposures In Unrestricted Unrestricted Areas from Maximum Hypothetical Accident Accident Assumed Assumed Failures Total Body Thyroid Thyroid Fraction of of Dose Dose Part 20 Annual Limits Fuel clad leak with normal operation of of 0.006 rem 0.0 10 rad 0.010 0.007 ventilation system; pool filled Fuel clad leak with failure of ventilation ventilation 0.084 rem 0.102 rad 0.018 0.018
- system; pool filled Fuel clad leak with failure of ventilation system and concurrent ventilation concurrent loss of pool water 0.153 rem 1.019 rad 1.019 0.052 0.052
- Safety Analysis UWNR Safety Report Rev. 2 Analysis Report 13-7 13-7 2008 Sept. 2008
13.1.2 13.1.2 Insertion of Excess Reactivity The worst case result of insertion of excess reactivity would be insertion of the maximum allowed experiment reactivity reactor is operating reactivity worth or ejection of the transient rod (1.4% operating at maximum maximum steady-state steady-state power. (1.4% LlKlK) AK/K) while the
- Calculations55 performed' Calculations performed by Gulf General Atomic indicate that a peak temperaturetemperature of 1150 'C°C in FLIP fuel will not produce a stress in the fuel clad in excess of the ultimate yield strength.
Further, TRIGA fuel with a H/Zr ratio of at least 1.65 has been pulsed to temperatures temperatures of about 1150 1150 °C'C without any damage damage to the clad 6.* In a mixed FLIP-Standard FLIP-Standard TRIGA TRIGA core the peak temperatures temperatures in FLIP fuel are much higher than in standard fuel due to the peaking of the power distribution near water gaps. For this reason the subsequent subsequent analysis in this section is concerned concerned with internal temperatures temperatures in FLIP fuel elements. A worst case core arrangement arrangement is considered, in which a FLIP element is located adjacent to a 3-inch square water gap. The power density in the FLIP element is at the maximum permissible maximum permissible value based on consideration consideration of the loss of coolant accident accident (23 KW when the core is operating at 1 MW). The core is operating operating at the power power level scram point of 1.25 MW, and the transient control rod is fired to initiate a pulse. Pulses of 2.1% AK/K fired in standard 2.1 % LlKlK standard TRIGA TRIGA at this facility have had energy releases of less than 20 MW seconds. FLIP and mixed cores cores have been operated with maximum reactivity
- insertions for pulses reduced to 1.4%LlKlK 1.4%AK/K because of the shortershorter prompt neutron lifetime in FLIP fueled cores. Typical 1.4% reactivity pulses in an all-FLIP Typical 1.4% all-FLIP core have energy releases releases of of only about 14 MW seconds, secohds, while mixed cores have slightly lower lower releases. Computations will Computations be done for 20 MW second second release.
The limitation of experiment reactivity to 1.4% 1.4% LlKlK AK/K will insure that reactivity insertions from experiment experiment removal or failure will insure that such an accident accident will result in consequences consequences no worse than those considered considered here. Firing the transient rod while at full power is prevented prevented by interlocks interlocks and administrative administrative requirements. Removal of an experiment while operating operating at full power would not result in a reactivity insertion rate as large as that resulting from firing the transient rod, and the most likely result of experiment experiment removal removal under under the conditions conditions assumed would be a reactor scram from power power level, and fuel temperature temperature trips. Further, experiments experiments having worths approaching 1.4% A approaching 1.4% Ll K/K KlK are fastened to prevent prevent inadvertent inadvertent removal, and administrative administrative restrictions do not allow such manipulations while the reactor reactor is in operation. The predicted predicted conditions establish an upper limit for a reactivity accident. UWNR Safety Analysis Report Rev. 2 13-8 13-8 Sept. 2008 2008
*
- 13.1.2.1 13 .1.2.l Fuel Temperatures Temperatures from Operation Operation at the Scram Point Calculations for the SAR SAR77 of the Puerto Rico Reactor resulted in the information presented presented in the lower curve in Figure 13-1 13-1. . This curve shows the fuel temperature temperature distribution at the axial centerline in a FLIP fuel element operating at conditions of slightly higher power power density than that assumed here. The Puerto Rico case is an element operating at a power density in the maximum elementelement of 1.4 times the average average of22.3 of 22.3 KW/element. The axial peaking factor is 1.3. Calculations 1.3. Calculations done for UWNR considered considered the case of an element operating at 23 KW times the ratio of 1.25 of scramscram setting/licensed power level, with the same axial peaking factor of 1.3. 1.3.
Using these numbers, the fuel centerline temperatures will be lower in the UWNR centerline and average temperatures UWNR core, but the temperature at the outer surface of the fuel would be approximately approximately the same in both cases. F,,. EMP. /A/ FtlEL TEMP. WY*X. 11/ MAX. POWER !)fNSITY POWER ELEM*EA/IT P[l=o)RE OVS/rY fLtMF.NT PE
..f1 AFTER ZO Mill A 20 AFTER ,4 stc.
Mw 5'Ec. ?tJLSE PULSE FRJM FRPJ/ /.25' M41 1/.2 Mw
.. AFKER /
- AFT[R PI)LS[
EfFOR[ 1'/)1..5£ Kz 200 f-I o~1--~--~~__~__~__~~__~__~~-/ t I t I P
'0 10 V030 ZJ 30 40 40 50 60 GO 70 70 80 80 90 90 /00 100
-1 OF 10 OTS/OE' FU'EL OF OtJTS//)E" RAOIIJS (%)
FVEL RAO/NJ (~~) 13-1 . Fuel Temperature
- Figure Figure 13-1 Temperature Distribution in a Fuel ElementElement UWNR UWNR Safety Analysis Analysis Report Rev. 2 13-9 13-9 2008 Sept. 2008
13.1.2.2 Temperature after Pulse Temperature Firing a pulse while at the scram point would cause the reactor to scram from power level and fuel temperature temperature scrams. The entire pulse energy release is used, however, in the following analysis.
- The temperature temperature distribution in the fuel element immediately immediately after a 20 MW MW second pulse is is plotted as the top curve in Figure 13-1. The peaking Figure 13-1. peaking factor within within a FLIP element element adjacent adjacent to a 3-inch square water-gap is 2.49, and an axial peaking peaking factor of 1.3 is used as in the steady state conditions. The energy deposited in the element under considerationconsideration is calculated calculated using the same peaking factor (power in maximum maximum element/
element/ power in average element in core) which resulted in the 23 KW steady state level. temperature reached in the element The maximum adiabatic temperature element will occur at the outer outer surface surface of the fuel element adjacent to the water-gap. This maximum temperature would be 1133 C, slightly below the safety limit of 1150 C. Although Although such an event is considered considered highly unlikely, it would not cause fuel damage or release release of fission products from the reactor. After After these computations computations were completed, NRC requested requested that the accident be re-evaluated re-evaluated for the permitted cores cores under the technical specifications technical specifications that were proposed proposed for UWNR. The major
- changes changes were limiting the minimum FLIP content to be 9 fuel bundles (35 elements, since since the transient transient rod is located within the 9 central fuel bundles in the UWNR UWNR core).
The re-analysis was basedbased on a lower power in maximally maximally exposed fuel element element (18.1 (18.1 kW instead of 22.3 kW) and a limitation of the reactivity insertion from 2.1 of22.3 % ~KlK 2.1% AK/K to 1.4% ~KlK. 1.4% AK/K. First, the temperature temperature of the fuel in the maximum element will be lower at the beginning of the
)
pulse by about 70°C. 70'C. Second, the use of a compact
.
compact array of nme nine (9) (9) FLIP bundles reduces the possible possible peaking factor within a FLIP element from the 2.49 value used in the original calculation to a value of 2.03 for a FLIP element of2.03 element beside the transient rod guide tube (this is the position position with highest power density in the core.)core.) Finally, reduction of allowable allowable pulsed reactivity insertion from 2.1% 2.1 % ~KlK AK/K to 1.4%1.4% ~KlK AK/K will substantially substantially reduce reduce the energy generation in a pulse, while the limitation of experiment worth to 1.4% 1.4% ~KlK AK/K will provide similar safeguards safeguards for experiment failure or removal. Measurements Measurements performed performed on the Puerto Puerto Rico Nuclear Center, Center TRIGA-FLIP reactor indicated that a pulse insertion of 1.4% AK/K insertion 1.4% t.KlK resultedresulted in a maximum fuel temperature rise of approximately 400 'C 8, approximately 400°C 8, and measurements confirmed that measurements at Wisconsin confirmed prediction. Consideration of all these differences temperature of about 450 'C differences shows a peak fuel temperature °C lower than that indicated above. It is therefore concluded that fuel damage would occur in neither neither case, but with a much larger safety margin in the more restrictive restrictive case considered here. UWNR Safety Analysis UWNR Analysis Report Report Rev. 2 13-10 13-10 2008 Sept. 2008
*
- 13.1.3 13.1.3 Loss of Coolant likelihood of complet~
Although there is little likelihood complete loss of water water from the reactor pool, an analysis is made to demonstrate that such loss will not damage reactor reactor fuel. 13.1.3.1 Possible Means Means of Water Loss The pool ~sis contained within the thick reinforced concrete concrete reactor reactor shield which will maintain its integrity under the most severe severe earthquake earthquake that would be expected expected in this area. The only credible credible scenario for draining the pool would be a sheared and open beam port. For For analysis in the next section, the time to drain the pool is estimated. The pool water level is routinely maintained maintained at least 20.875ft 20.875ft above the top of the core (this is approximately the low pool level alarm point) which corresponds corresponds to 21.5ft above the fuel center. However, the limiting assumption is made that the water level is only 19ft above the top of the core, or 19.625ft 19.625ft above the fuel center. The pool has a surface area of 89. 89.13ft 2 13ft2.. The inner beam port diameter is 0.5ft, and the beam ports are at core center. Using these values, the following equation equation99 can be used to estimate the time required required to drain the pool.
- where: tdtd =
Ap
= time to drain pool to height h (sec) g == acceleration Ap == cross-sectional (sec) 174ft1s 2))
(32.174ft/s acceleration due to gravity (32. cross-sectional area of pool surface Ao == cross-sectional surface (ft 2) (fe) cross-sectional area of drain opening (ft2)(ft2) Cd = = discharge coefficient (0.6) h= ho = initial height of water above drain opening (ft) opening (ft) h = final height of water above drain opening (ft) The calculated drain time is 836s. A shearedsheared and open beam port could drain the water level to mid-core mid-core height, but water would still be in contact with the fuel and would prevent excessive temperatures. temperatures. The 8-inch stainless stainless steel pipes built into the pool walls for possible future use in a forced convection cooling system are flange sealed on the outer ends. In addition, one of these pipes has a loop and a siphon breaker breaker extending extending well above the core so that a rupture cannotcannot lower pool level below the core. The other pipe is flange sealed inside the pool and penetratespenetrates the shield shield wall well above the core. Rupture Rupture of either of these lines will not uncover uncover the core. Rupture of the piping in the dernineralizer demineralizer could cause only slight water loss due to location location of of the outlet lines from the pool and a check check valve at the demineralizer demineralizer outlet.
- UWNR UWNR Safety Analysis Report Rev. 2 13-11 13-11 2008 Sept. 2008
I I 13.1.3.2 1 13.1.3.2 Calculations Radiation Radiation Levels in Confinement Calculations of radiation levels 'at continuous operations at continuous operations 1.02MW. the ORIGEN2 1.02MW. The computer code version 2.1, ORIGEN2 computer fission product product Unshielded Core Confinement Due to Unshielded at various points in the Reactor Reactor Laboratory inventory Laboratory were made assuming was determined 2.1, using the PWRUS cross-section determined by the use ofof cross-section library. A single fuel
- I pin was simulated simulated at the power power level of the hottest rod, with an assumed exposure of 500MWd.
500MWd. I The resulting gamma source term was multiplied by the number of fuel elements in the core and I then divided by the pin power peaking factor to calculate calculate the core gamma source term at various I decay decay times. The decay time of 836s represents represents the time required to drain the pool (see section section I 13.1.3.1). Dose rates from direct and scattered radiation were modeled using the MCNP5 13.1.3.l). I computer computer code. Results of the calculations are given in Table 13.3. 13.3. Table 13.3 Table 13.3 Calculated Calculated Radiation Radiation Dose Rates in Confinement Confinement After Pool Water is Lost Lost Time After Beam Port Console Pool Curb Pool Top Shutdown Floor Over Core Behind Curb (R/hr) (Rlhr) (R/hr) (Rlhr) (R/hr) . (R/hr) (Rlhr) 836 seconds 4.89 6.72 7,430 28.1 1 day 1.23 1.69 1,720 1,720 7.07 1 week 0.682 0.938 992 3.92 1 month 0.353 0.485 492 These levels are not too high to allow emergency repairs to be made. Facility procedures cover the situation of pool water loss. procedures 2.02 Facility emergency
- 1 13.1.3.3 13.1.3.3 Radiation Radiation Levels in Unrestricted Unrestricted Areas Due to Unshielded Unshielded Core The calculations from the previous section were also made for the 33r" rd floor non-restricted non-restricted classroom to the west of the reactor. This classroom would be subject to the highest dose rate field of any non-restricted non-restricted area due to the elevation above the biological biological shield and its line-of-sight with the reactor core.
UWNR Safety Analysis Report Rev. 2 13-12 13-12 2008 Sept. 2008
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- 13.4 Table 13.4 Calculated Radiation Dose Rates in Non-Restricted Non-Restricted Area After Pool Water is Lost Time After 3 rd Floor Classroom 3'd Classroom Shutdown (R/hr)
(Rlhr) 836 seconds 4.14 4.14 1 day 0.764 1 week 0.482 1 month 0.210 0.210 The calculated calculated dose rate to the 33rdrd floor non-restricted non-restricted classroom is significant, but in the event of of a loss of coolant accident the building evacuation evacuation alarm would alert people to evacuate these classrooms before the core was completely completely uncovered. In order to estimate the integrated dose received by a member of the public during the evacuation, the MCNP5 model of the unshielded received core was modified to include include partial water shielding at several time steps. The core gamma gamma source term was also modified to simulate an appropriate level of decay from full power. The integrated dose to the 33 rdrd floor classroom classroom was calculated calculated at various times during the pool water loss and is shown in Figure 13-2. 13-2. 250 250 ------,
- Core i 200 uncovered uncovered 7 \
I E 0
...
(I) I..
.§.
0 (I) 150 150 /
/
I/) 00 building 5 minute building 00
-
"0 0 (I)
.6~
...c(UI..u 100 evacuation complete evacuation complete
/
I
- 0)
CI 0 (I)
.4-.
c: 50 Bridge ARM ARM initiates evacuation initiates evacuation alarm
,!
,i
!
I 50 o0 i V I i I i o0 100 200 300 400 500 600 700 800 90 900 )0 Time Since Since Start of LOCA LOCA (sec) (sec) Figure 13-2 13-2 Integrated Dose vs. Time during LOCA 33 rdrd Floor Integrated LOCA
- UWNR Safety Analysis Report Rev. 2 13-13 13-13 2008 Sept. 2008
A 5 minute evacuation time from drain to approximately monitor, which would in tum from the sounding of the evacuation alarm is assumed. If the initiation of the pool water loss was observed then the building evacuation would be manually activated at the start of the LOCA, but if the accident was unobserved then the pool water would activated approximately 7.4 ft above the core in 300 s before tripping the bridge area radiation turn automatically automatically initiate the building evacuation radiation evacuation alarm. Therefore, Therefore, the
- hypothetical hypothetical member ofthe rd floor classroom for 5 minutes following of the public that remains in the 3yd evacuation alarm (300 s after start of the LOCA) would automatic initiation of the building evacuation the automatic receive an integrated dose of about 13 mrem. This is less than the 100mrem 100mrem limit (10 (10 CFR CFR 20.1301). Realistic doses would be far less than this, because because the preceding analysis does not stairwells (where the dose rate is much lower) take into account time spent in hallways and stairwells during the evacuation. Because the time spent in the high dose rate field in the 3yd rd floor classroom classroom would be far less than 5 minutes, the integrated integrated dose would be substantially lower due I to the majority of the dose being received in the final minute as shown in Figure 13-2. 13-2.
13.1.3.4 13.1.3.4 Fuel Temperature Temperature After Loss of Pool Water Calculations Calculations performed performed at Texas A & & M University have treated the loss of coolant accident in detail, based on reactor shutdown 15 minutes beforebefore the core is uncovered. At Wisconsin, the pool level scram would cause automatic automatic shutdown much sooner, as the A & & M calculation is based on pool drainage by rupture of a 10-inch based 'on 10-inch line. Other pertinent parameters of the two pertinent parameters facilities are identical. The calculations employed the Gulf computer computer code TAC for calculation calculation
- of system temperatures.
temperatures. The results of these calculations calculations (Pages 25-31 of Texas Texas A & & M University University Nuclear Nuclear Science Center Amendment Amendment II to the Safety Analysis Report, November 1, 1, 1972 submitted submitted under Docket for License License R-83) indicate that for a maximum power of less than 21 kW/element power density ofless kWlel,ement for standard fuel and 23 kW/element k Wlelement for FLIP fuel, loss of coolant water would not result in fuel clad failure and release release of fission products. products. 13.1.4 Loss of Coolant FlowFlow Not applicable; natural convection cooling natural convection Mishandling or Malfunction 13.1.5 Mishandling Malfunction of Fuel Reference 1 states that this condition produces the maximum consequence Reference consequence to the public. This accident is therefore included included as the maximum accident (Section 13.1/2.1), when maximum hypothetical accident combined with failure of the ventilation system and loss of pool water. The effect of a fuel clad failure with normal pool level and with the ventilation ventilation system operating operating normally has no significant significant effect on the public. UWNR UWNR Safety Safety Analysis Report Report Rev. 2 13-14 13-14 2008 Sept. 2008
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- 13.1.6 Experiment 13.1.6 Malfunction Experiment Malfunction Experiment composition are controlled Experiment reactivity worth and composition controlled and and limited so that experiment experiment failure will not insert a step step change in reactivity reactivity greater 1.4% LlKlK greater than 1.4% AK/K (fixed experiments; experiments; experiments are limited to 0.7%
movable experiments 0.7% AK/K. Procedure for experiment LlKlK. Procedure experiment review review includes includes consideration of chemical and explosive consideration explosive hazards hazards to the reactor. Any experiment experiment containing containing fissionable material is limited so that production fissionable production of gaseous products results gaseous and volatile fission products in releases lower lower than that considered 13. 1.1. Therefore, considered in section 13.1.1. Therefore, experiment experiment malfunction malfunction will not result in consequences consequences moremore severe severe than those listed in other parts of this chapter. chapter. 13.1.7 Loss of Normal Electrical Power 13.1.7 Power Loss of normal electrical power will cause cause the reactor reactor to shut down. It will not result in any release release of radioactive material or increase increase the dose to the population. Emergency Emergency corecore cooling cooling engineered engineered safety systems systems are not required. The maximum hypothetical maximum hypothetical accident analysis does analysis include include loss of the ventilation ventilation system in the analysis, thus effectively electrical effectively including loss of electrical power. 13.1 .8 External Events 13.1.8 Since the safety of a TRIGA TRIGA reactor reactor is so strongly a function function of the fuel composition composition and
- characteristics, none characteristics, external event none of the usual external event initiators initiators will cause any effect on the public. As stated in chapter 2 of this report, floods and hurricanes hurricanes are an insignificant insignificant threat threat to the safety safety of of (Section 2.4), with the 100 the reactor (Section 100 year flood causing nearby nearby Lake Mendota Mendota to expand expand byby only 3030 feet, not threatening threatening the laboratory.
laboratory. Further, should should the water level level within within the room riserise above ground level it would not affect affect the safety safety of the reactor. Tornados Tornados do occur occur in the concrete shield Midwest, but damage to the concrete shield which protects protects the reactor core is not credible. credible. The seismicity seismicity of the area area is extremely estimated 50 extremely low, with the estimated 50 year peak ground acceleration to a ground acceleration seismic event seismic 0.01 gg (Section event is less than 0.01 (Section 2.5.4). Since structures other than the engineered structures Since no engineered reactor required to provide reactor shield are required reactor, such an acceleration provide protection to the reactor, acceleration will have have no effect reactor. Likewise, though aircraft collisions effect on the reactor. collisions with the building are not impossible, impossible, they are unlikely due to the location location of impacts will not breach of flight paths. Further, such impacts breach the concrete shield at core level. Impacting the outer concrete walls of the building will will not result in radiation radiation being released.. being released
- UWNR Safety Analysis Report UWNR Report Rev. 2 13-15 1-5Sp.20 13-15 2008 Sept. 2008
e Malfunction of Equipment 13.1.9 Mishandling or Malfunction An analysis analysis was made of the possibility possibility that loss of water from the reactor or from the radioactive liquid waste storage tanks could affect the city water supply, and a negative result result was obtained as indicated below. Madison obtains its drinking drinking water supply from several several wells drilled into the Cambrian Cambrian sandstone described described above. The location of these wells is shown onon Figure 2-12, and they supply the University University as well as the city. All of these wells are cased from . ground level into the sandstone so as to keep out water from the glacial deposit. The closest well to the reactor site is about 2,000 feet southeast. The Reactor Laboratory Laboratory floor drain empties into the hold tank. Should the entire contents of the pool be let out into the room, however, some water could escape into the sewer system through a drain thimble into which which waste water is pumped pumped from the hold tank. There are four methods by e. which water may leave the reactor room: (1) (1 ) by flow pumped from the radioactive waste storage tanks through elevated through the elevated drain thimble provided for emptying the tanks; (2) (2) directly into the drain drain thimble should the pool be completely ruptured, thus reaching reaching a level high enough to overflow into the thimbles or escape from the laboratory and enter floor drains in surrounding areas; (3) by loss through the floor and into the ground; and (4) by rupture of the liquid radioactive waste storage tank directly into the soil under the laboratory floor. Analysis for cases (1) and (2)(2) In so far as the first two discharge paths are concerned, concerned, the flow through the drain thimble or floor drains empties into a sanitary sewer main. From there it would travel through mains via a pumping station to the main sewage plant, located south south of and outside the corporate limits of the city. From there, the sewage travels through mains an additional five miles to the south before it empties into an open ditch. On the way, any UWNR Safety Analysis Report Rev. 2 13-16 13-16 Sept. 2008 e
- considerably diluted since the minimum flow-rate water from the reactor would become considerably into the ditch is 7,000 gpm whereas the probable maximummaximum rate of entry into the floor drain would not be more than 10 to 100 gpm. These facts, coupled with the fact that stringent administrative administrative precautions will be taken to ensure that water contaminated beyond established established tolerance contaminated tolerance levels is not released to the drain, tend to preclude city water supply could be adversely adversely affected by this method.
preclude that the Analysis for case (3) The possibility that the city water supply could be affected via the third method (3) is also negligible. negligible. The base of the reactor is about 8 feet below ground level, and water cannot cannot be dissipated via surface surface run-off. Since Since the walls ofthe of the building surrounding reactor surrounding the reactor are made of concrete concrete up to the ground level, significant water loss through the floor could could result only if the concrete was breached. In fact, it would appear that the only mechanism by which contaminated contaminated water could enter the soil would be the result of an earthquake sufficiently severe to rupture both the reactor tank and shield, as well as the floor of the sufficiently building, at a time when the reactor pool water was radioactive beyond tolerance levels. coincidental occurrences Such a set of coincidental occurrences is considered extremely remote. Further, even if it did occur, there is no assurance assurance that the water supply would be adversely affected. For example, example, the nearest city well is about 2,000 feet from the reactor reactor site, and it has been been estimated estimated by a ground water specialist that water would flow through the sandstone sandstone from from the reactor to the well at not more than 0.1 foot per day. Thus, as long as 55 years might
- be required required for the reactor reactor water to reach the well.
Analysis for case (4) Should the radioactive waste storage tanks rupture, a similar analysis to that in case (3) indicates no adverse adverse effect on the well. Furthermore, Furthermore, the quantities of water likely to be lost are small and activities activities are expected to be low enough that no hazard exists. 13.2 13.2 Summary and Conclusions None of the accidents considered here will result in consequences accidents considered consequences to the public health and safety. Even the maximum hypothetical accidentaccident does not result in releases of radioactivity in excess of of 10 CFR Part 20 limits when averaged over a year. 13.3 13.3 References 1.
- 1. NUREG- 1537 Part 1, NUREG-1537 1, Guidelines Guidelines for Preparing and Reviewing Applications Applications for the Licensing of Non-Power Reactors, Licensing Reactors, USNRC, February 1996. 1996.
- 2. Credible Accident Analyses for TRIGA and TRIGA-fueled Credible Accident TRIGA-fueled Reactors, NUREG/CR-2387, NUREG/CR-2387, PNL-4028, PNL-4028, April 1982 1982..
- UWNR Safety Analysis Report Rev. 2 13-17 13-17 2008 Sept. 2008
3.
- 4. COMPLY COMPL R. W. King, "Energy J. F. Perkins and R.
Nuclear Science and "Energy Release from the Decay of Fission Products", and Engineering, Engineering, 3, 726 (1958). Y V1.5d, U.S. Environmental (1958). Products", Environmental Protection Agency, Office of Radiation and Indoor air, Washington DC 20460, October 1989.
- 5.
- 5. Reactor", GA-9064, Gulf "Safety Analysis Report for the Torrey Pines TRIGA Mark III Reactor", Gulf General Atomic, Jan. 5,5, 1970.
- 6. "Annular Core Pulse Reactor", General Dynamics, General Atomic Division Report "Annular 2, 9/30/66.
GACD 6977, Supplement 2,9/30/66. c
- 7. Safeguards Summary Report for the TRIGA-FLIP Reactor at Puerto Rico Nuclear Safeguards Nuclear Center, Report PRNC 123, Revision C, November 11, 11, 1969.
8.
- 8. Technical Specifications Docket 50-120, Change No. 11 to the Technical R-83, Specifications Facility License R-83, Texas A && M University, Section Section 3.2 Basis.
- 9. Hunsaker and Rightmire, "Engineering "Engineering Applications of Fluid Mechanics,"
Mechanics," McGraw-Hill McGraw-Hill Book Company, 1947 (pp. 69-70). 0
- UWNR Safety Analysis UWNR Safety Analysis Report Rev. 2 13-18 13-18 Sept. 2008 2008
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- 14 TECHNICAL TS INTRODUCTION TS 1.1 TS 1.1 Scope SPECIFICATIONS TECHNICAL SPECIFICATIONS TS 11 INTRODUCTION This section of the SAR for license renewal of the University University of Wisconsin Nuclear Wisconsin Nuclear Reactor Reactor constitutes the proposed proposed Technical Specifications for that facility as required Technical Specifications required by 10 CFR 50.36. This document document includes includes the basis to support support the selection and significance of the specifications. information purposes specifications. Each basis is included for information only, and is not part of the Technical Specifications in that it does not constitute Technical Specifications requirements requirements or limitations which the licensee must meet in order to meet meet the specifications. Dimensions, measurements, and other numerical values given in these specifications.
specifications may differ slightly from actual values due to construction and specifications manufacturing manufacturing tolerances or normal degree degree of accuracy or of instrument readings. specifications are re-formatted These specifications re-formatted from the technical specifications specifications in force in 1999. Changes changes required by name changes or to include Changes reflect only changes information information not in the original original technical specifications. In addition, certain additions required by NUREG-1537 NUREG-1537 are included. All substantive substantive changes were denoted by redlining redlining in Rev 0, but currently only changes changes since the last revision are redlined
(indicated by vertical line in margin). These technical specifications specifications continue to TRIGA-FLIP and the original LEU TRIGA fuels, either separately include use of TRIGA-FLIP separately or in mixed cores. TS TS 1.2 Format Content and section numbering numbering is in accordance accordance with section 1.2.2 of ANSI/ANS ANSI!ANS 15.1. 15.1. TS 1.3 TS 1.3 Definitions The terms used herein are explicitly defined to ensure uniform interpretation of the uniform interpretation Technical Specifications. Technical Specifications .
- UWNR Safety Analysis Report Rev. 2 14-1 14-1 2008 Sept. 2008
TS 1.3.1 TS 1.3.1 Reactor Operating COLD Operating Conditions COLD CRITICAL: CRITICAL: The reactor reactor is in the cold critical condition when it is critical with the fuel and bulk water temperatures temperatures both below 12soF. 125'F.
- PULSE MODE (PU)
Pulse mode operation shall mean mean any operation operation of the reactor with the ( mode selector selector switch in the pulse position. REACTOR REACTOR SECURED: The reactor is secured secured when:
- 1. Either there is insufficient moderator available
- 1. available in the reactor to attain criticality or there is insufficient insufficient fissile material present in the reactor to attain criticality upon optimum available conditions conditions of moderation moderation and reflection, or
- 2. The following conditions exist:
- a. The reactor reactor is shut down,
- b. The console key switch is.in the "off' position and the key is removed from the console and under the control of a licensed operator operator or stored in a locked storage area, and
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- c. No work is in progress progress involving in-core fuel handling or refueling operations, maintenance of the reactor or its control control mechanisms, or insertion or withdrawal of in-core experiments with a reactivity exceeding 0.7% ~KlK.
worth exceeding AK/K. REACTOR REACTOR SHUTDOWN: The reactor is shut down when the reactor is subcritical subcritical by least 0.7% ~k!k Ak/k of reactivity. REACTOR OPERATION: operation is any condition Reactor operation condition wherein wherein the reactor is not secured. UWNR UWNR Safety Analysis Report Report Rev. 2 14-2 14-2 2008 Sept. 2008
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- REPORTABLE REPORTABLE OCCURRENCE:
OCCURRENCE: A reportable occurrence occurrence is any of the following that occur during reactor reactor operation:
- 1. Operation with any safety system setting less conservative
- 1. conservative than specified in the technical specified technical specifications; specifications;
- 2. Operation in violation of a Limiting Condition for Operation Operation listed in Section 3;
- 3. Operation with a required reactor or experiment experiment safety system system component component in an inoperative inoperative or failed condition which could render render the system incapable incapable of performing performing its intended safety function;
- 4. Any unanticipated unanticipated or uncontrolled change in reactivity reactivity greater than AK/K, excluding 0.7% LlKlK, excluding reactor trips from a known cause;
- 5. An observed inadequacy
- 5. inadequacy in the implementation implementation of either administrative or procedural controls, such that the inadequacy administrative inadequacy could have caused caused the existence development of a condition which could existence or development result in operation of the reactor outside the specified safety limits; and
- 6. Abnormal Abnormal and significant degradation degradation in reactor fuel or cladding which which
- SHUTDOWN could result in exceeding prescribed radiation personnel personnel or environment, or both.
SHUTDOWN MARGIN:* MARGIN: radiation exposure exposure limits of of Shutdown margin shall mean the minimum shutdown reactivity necessary to provide provide confidence confidence that the reactor can be made subcritical subcritical by means ofof the control and safety systems, starting. from any permissible operating condition condition (assuming the most reactive scrammable scrammable control element and any non-scrammable non-scrammable control elements remain full out), and the reactor will remain subcritical subcritical without further operator action. SQUARE WAVE MODE (SW) SQUARE (SW) Square Square wave mode operation operation shall mean any operation of the reactor with the mode selector selector switch in the square wave position. STEADY STATE MODE (SS) (SS) Steady state mode operation shall mean operation of the reactor with the mode selector selector switch in the manual or automatic positions positions..
- UWNR UWNR Safety Analysis Report Report Rev. 2 14-3 14-3 Sept. 2008 2008
TS TS 1.3.2 1.3.2 Reactor Experiments and Irradiation Reactor Experiments
. EXPERIMENT:
Experiment shall mean: Irradiation
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- 1. Any apparatus, device or material which is not a normal
- 1. normal part of the reactor core or experimental experimental facility, or
- 2. Any activity external to the biological biological shield using a beam of radiation radiation emanating from the reactor emanating reactor core, or
- 3. Any operation operation designed designed to measure reactor parameters or or characteristics, characteristics, or any activity external to the biological biological shield using a beam of radiation emanating from the reactor core:
Classification experiments shall be: Classification of experiments
- 1. Routine experiments.
- 1. experiments. Routine experiments experiments are those which have previously been been performed at the facility.
- 2. Modified experiments. Modified Modified routine experiments. experiments are those Modified routine experiments which have not been performed previously but are similar to the
- routine routine experiments experiments in that the hazards are neither greater nor significantly different than those for the corresponding corresponding routine experiments.
- 3. Special experiments. Special experiments are those which which are not routine routine or modified experiments.
EXPERIMENTAL FACILITIES: EXPERIMENTAL Experimental facilities shall mean beam ports, including extension Experimental extension tubes with shields, thermal columns with shields, vertical tubes, through tubes, in-core irradiation baskets, irradiation cell, pneumatic transfer systems and any other in-pool irradiation irradiation facilities. IRRADIATION: IRRADIATION: Irradiation Irradiation shall mean the insertion of any device or material that is not a normal part of the core or experimental experimental facilities into an experimental experimental facility so that the device.or material device or material is exposed to a significant amount of of the radiation available in that irradiation irradiation facility. UWNR Safety Analysis Analysis Report Report Rev. 2 14-4 14-4 2008 Sept. 2008
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- NON-SECURED EXPERIMENT NON-SECURED Any experiment SECURED EXPERIMENT:
SECURED EXPERIMENT experiment not meeting the criteria EXPERIMENT: criteria of a secured secured experiment. experiment. A secured A secured experiment experiment shall mean mean any experiment experiment that is held firmly inin place place byby a mechanical device or by mechanical device by gravity, gravity, that is not readily removable removable from the reactor, and that requires one of the the following following actions to permit permit removal:
- 1. Removal
- 1. Removal of mechanical mechanical fasteners
- 2. Use of underwater underwater handling handling tools 3.
- 3. Moving Moving of shield blocks or beam port containers.
containers. TS 1.3.3 TS 1.3.3 Reactor Components Components LATTICE POSITION: CORE LATTICE POSITION: A core lattice position is that region in the core (approximately A (approximately 3"3" by by 3") 3) over a grid hole. It may be by a fuel bundle,
- occupied by be occupied bundle, an experiment experiment oror experimental facility, or a reflector experimental reflector element.
FUEL BUNDLE: FUEL BUNDLE: A A fuel bundle bundle is a cluster cluster of three or four fuel elements elements secured secured in a square square array byby a top handle handle and a bottom bottom grid plate adaptor. FUEL ELEMENT: FUEL ELEMENT: A fuel element is a single TRIGA fuel rod of either standard A standard or FLIP type. FLIP CORE: A FLIP core A arrangement ofTRIGA-FLIP core is an arrangement of TRIGA-FLIP fuel in the reactor gridgrid plate. plate. FLIP FUEL: FUEL: FLIP fuel is TRIGA TRIGA fuel that contains contains a nominal 8.5 weight nominal 8.5 weight percent of percent of uranium with a 11 5 235UU enrichment enrichment of about 70%70% and erbium erbium as burnable burnable poison.. pOlson
- UWNR UWNR Safety Analysis Report Rev. 2 Safety Analysis 1-5Sp.20 14-5 14-5 2008 Sept. 2008
INSTRUMENTED ELEMENT: INSTRUMENTED An instrumented instrumented element is a special fuel element in which thermocouples are embedded embedded for the purpose of measuring reactor reactor operation. measuring fuel temperatures thennocouples temperatures during
- MIXED MIXED CORE:
A mixed core is an arrangement arrangement of standard TRIGA fuel elements and and FLIP fuel elements elements with at least 35 TRIGA-FLIP fuel elements located in a central region of the core. OPERATIONAL CORE: OPERATIONAL An operational operational core may be a standard core, mixed core, or FLIP core for which the core parameters parameters of shutdown margin, fuel temperature, temperature, power power allowable reactivity calibration, and maximum allowable reactivity insertion insertion have been determined to satisfy the requirements of the Technical requirements ofthe Technical Specifications. REGULATING BLADE: REGULATING The regulating blade is a low worth control blade that need not have scram scram
- capability. Its position may be varied manually or by the servo-controller.
SHIM-SAFETY SHIM-SAFETY BLADE: A shim-safety blade is a control blade having an electric motor drive andand scram scram capabilities. Its position may be varied manually or by the servo-controller. STANDARD TRIGA TRIGA FUEL: Standard TRIGA fuel is TRIGA fuel that contains a nominal 8.5 weight percent 235U enrichment percent of uranium uranium with a 235U enrichment of less than 20% 20% and no burnable burnable poison. STANDARD CORE: A standard standard core is an arrangement of standard standard TRIGA TRIGA fuel in the reactor reactor grid plate. UWNR UWNR Safety Analysis Report Rev. 2 14-6 14-6 2008 Sept. 2008
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- TRANSIENT TRANSIENT ROD:
The transient rod is a control rod with scram capabilities that can be rapidly ejected ejected from the reactor solid aluminum aluminum follower. reactor core to produce a pulse. Its position may be varied manually or by the servo-controller. servo-controller. It may have a voided or TS 1.3.4 TS Reactor Instrumentation: Reactor CHANNEL CHANNEL CALIBRATION: CALIBRATION: A channel channel calibration calibration consists consists of comparing comparing a measured value from the measuring channel measuring channel with a corresponding known value of the parameter corresponding parameter soso that the measuring channel output can be adjusted adjusted to respond with acceptable accuracy accuracy to known values of the measured variable. variable. CHANNEL CHANNEL CHECK: A channel check is a qualitative verification verification of acceptable performance by acceptable performance observation of channel behavior. observation
- CHANNEL TEST:
CHANNEL A channel test is the introduction introduction of a signal signal into the channel to verify verify that it is operable. operable. EXPERIMENT EXPERIMENT SAFETY SAFETY SYSTEMS: SYSTEMS: Experiment Experiment safety systems systems are those systems, including including their associated input circuits, which are designed to initiate a scram for the primary purpose purpose of protecting protecting an experiment or to provide information whichwhich requires requires manual protective action to be initiated. LIMITING LIMITING SAFETY SAFETY SYSTEM SETTINGS: SETTINGS: Limiting safety system settings are settings for automatic protective devices related to those variables having significant safety functions. MEASURED VALUE: MEASURED The measured value is the magnitude of that variable variable as it appears on the output of a measuring channel.
- UWNR UWNR Safety Analysis Report Rev. 2 14-7 14-7 2008 Sept. 2008
MEASURING MEASURING CHANNEL: A measuring channel is the combination or lines, amplifiers, combination of sensor, interconnecting cables amplifiers, and output device of measuring cables connected for the purpose device which are connected measuring the value of a variable.
- OPERABLE:
OPERABLE: A system, device, or component component shall be considered operable when it is performing its intended functions ih capable of performing in a normal manner. REACTOR SAFETY SAFETY SYSTEMS: SYSTEMS: Reactor safety systems are those systems, including including their associated input circuits, which are designed to initiate a reactor scram for the primary purpose purpose of protecting protecting the reactor or to provide information which requires manual protective protective action to be initiated. SAFETY SAFETY CHANNEL: CHANNEL: A safety channel is a measuring measuring channel channel in the reactor safety system. SAFETY LIMITS: LIMITS: Safety limits are limits on important barriers which guard important process against the process variables which are found to be necessary to reasonably protect the integrity of certain barriers uncontrolled uncontrolled certain of the physical release of radioactivity.
- UWNR UWNR Safety Analysis Analysis Report Rev. 2 14-8 14-8 2008 Sept. 2008
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- TS 2 TS TS 2.1 Safety TS LIMITS AND SAFETY LIMITS Safety Limits Applicability Applicability LIMITING SAFETY AND LIMITING SAFETY SYSTEM SETTINGS SYSTEM SETTINGS This specification specification applies to fuel element element temperature and steady-state reactor reactor power level.
Objective The objective objective is to define the maximum fuel element temperature temperature and reactor power level that can be permitted with confidence that no fuel element element cladding cladding failure will result. Specifications
- 1. The temperature TRIGA-FLIP fuel element shall not exceed 1150°C temperature in a TRIGA-FLIP 1150'C under any conditions of operation.
- 2. The temperature temperature of a standard TRIGA fuel element element shall not exceed exceed 1000°C under under
'. any conditions of operation.
- 3. The reactor steady-state power reactor steady-state exceed 1500 kW under any power level shall not exceed conditions of operation.
Basis A loss of integrity of the fuel element cladding could arise from a buildup of of excessiye pressure between the fuel moderator and the cladding if the fuel excessive temperature exceeds exceeds the safety limit. The pressure is causedcaused by air, fission product gases, and hydrogen hydrogen from dissociation dissociation of the fuel moderator. The magnitude magnitude of this; this: pressure is determined determined by the fuel moderator temperature temperature and the ratio of hydrogen to zirconium zirconium in the alloy. The safety limit for the TRIGA-FLIP TRIGA-FLIP fuel element is based based on data which indicate indicate that the stress in the cladding cladding due to hydrogen pressure pressure from the dissociation of zirconium hydride hydride will remain below the ultimate stress provided the temperature does not 0 1150 C and the fuel cladding cooled'.1. exceed 1150°C cladding is water cooled The safety limit for the standard standard TRIGA fuel is based on data including the large amount of experimental experimental evidence obtained during high performance performance reactor tests ofof this fuel. These data indicate that the stress in the cladding (due to hydrogen pressure
- UWNR Safety Safety Analysis Report Rev. 2 14-9 14-9 2008 Sept. 2008
from the dissociation dissociation of zirconium hydride) will remain below the ultimate stress provided that the temperature temperature of the fuel does not exceed 1000 cladding is water cooled It has been shown cooled'.2
- shown by experience 100000 eC and the fuel experience that operation of TRIGA reactors at a power level of of
.'
1500 kW will not result in damage to the fuel. Several reactors of this type have operated successfully operated successfully for several years at power levels up to 1500kW. 1500kW. It has beenbeen shown ~nalysis and by measurements shown by analysis measurements on other TRIGA reactorsreactors that a power power level of 1500 kW corresponds corresponds to a peak fuel temperature temperature of approximately 600 oe. Thus a approximately 600'C. Safety Limit on power power level of 1500 1500 kWk W provides an ample margin of safety for operation. TS 2.2 Limiting Safety System Settings TS Applicability Applicability specification applies to the scram setting This specification setting which prevents the safety safety limit from from being reached. Objective The objective objective is to prevent Specifications
- 1. The limiting safety prevent the safety limits from being being reached.
safety system setting for fuel temperature temperature shall be 400'C 400 0 e (750'F) (750°F) as
- measured measured in an instrumented instrumented fuel element. For Foraa mixed core, the instrumented instrumented located in the region of the core containing element shall be located element containing FLIP type elements.
- 2. The limiting li.!lliting safety system setting for reactor power level shall be 1.25 MW.
Basis The first limiting safety temperature which, if exceeded, shall safety system setting is a temperature cause a reactor scram scram to be initiated initiated preventing preventing the safety safety limit from being exceeded. A setting of 400'C 400 0 e provides provides a safety margin of750oe of 750'C for FLIP type fuel elements elements and a margin of600oe of 600'C for standard TRIGA fuel elements. A part of the safety margin is used to account for the difference difference between between the true and measured temperatures temperatures resulting from the actual location of the thermocouple. If the thermocouple element is located located in the hottest position in the core, the difference difference between the true and temperatures will be only a few degrees measured temperatures degrees since the thermocouple thermocouple junction junction is at the mid-plane of the element element and close to the anticipated hot spot. If the UWNR UWNR Safety Safety Analysis Analysis Report Report Rev. 2 14-10 14-10 2008 Sept. 2008
*
(
"
- thermocouple element thermocouple element is located in a region of lower temperature, periphery of the core, the measured periphery that actually occurring measured temperature temperature, such as on the temperature will differ by a greater amount from occurring at the core hot spot. Calculations and measurements the facility with all permitted mixed core arrangements true temperature temperature at the hottest location temperature measurements made at arrangements indicate that, for this case, the location in the core will differ from the measured temperature by no more than a factor of two. Thus, when the temperature temperature in the thermocouple thermocouple elements reaches the trip setting of 400°C, 400'C, the true temperature temperature at the hottest location would be no greater than 800°C 800'C providing a margin to the safety limitlimit of at least 200'C 200°C for standard standard fuel elements and 350'C 350°C for FLIP type elements. These margins are ample to account for the remaining remaining uncertainty in the accuracy accuracy of the fuel temperature measurement channel and any overshoot measurement channel overshoot in reactor power resulting from a reactor transient during steady steady state mode operation. For a mixed core (i.e., one containing both standard and FLIP type elements),
containing elements), the requirement requirement that the instrumented element be located in the FLIP region of the core provides an even instrumented greater greater margin of safety since the peak to average power ratio within that region will be smaller than over an entire core composed of elements of the same type. Calculations and measurements measurements for this and similar TRIGA TRIGA reactors indicate at 1.25 MW, the peak fuel temperature temperature in the most limiting core loading permitted permitted under specifications (9 FLIP bundles) will be less than 600'C section 3 of these specifications 600°C so that the second limiting power level setting provides an ample safety margin to accommodate accommodate
- errors in power level measurement measurement and anticipated operational transients.
anticipated operational transients . In the pulse mode of operation, the first limiting safety system setting will apply. However, the power power level channels do not provide protection protection in pulse mode, and the temperature channel channel will have no effect on limiting the peak powers generated because because of its relatively relatively long time constant (seconds) (seconds) as compared compared with the width of of the pulse (milliseconds). (milliseconds). The limit on transient rod worth in another specification another specification limits the generated power so that fuel temperatures temperatures reached in a transient transient are smaller than those from full power operation. This transient rod worth limit is less than the reactivity required required for steady-state If the transient rod fails to steady-state full power. Ifthe automatically automatically drop after the pulse, fuel temperature reached due to energy generation generation in the tail of a pulse will be less than that at full power. Only in the case of operation outside the permitted parameters of core composition and reactivity reactivity limitations would fuel temperature temperature safety system actuation be needed to provide protection protection in any mode.. operating mode
- UWNRSafty nalsis epot Rv.
UWNR Safety Analysis U"R Analysis Report Report Rev. 2 14-1 14-11 14-11 Spt.200 Sept. 2008 2008
- This page is intentionally left blank.
- UWNR Analysis Report Safety Analysis UWNR Safety Report Rev. 2 14-12 14-12 2008 Sept. 2008
*
- TS 33 TS LIMITING CONDITIONS LIMITING TS 3.1 TS CONDITIONS FOR OPERATION 3.1 Reactor Core Parameters Applicability Applicability OPERATION These specifications specifications apply to the reactivity condition of the reactor reactor and the reactivity worths of control rods. They apply for all modes of operation.
Objective The objective objective is to assure that the reactor can be shut down at all times and to assure that the fuel temperature temperature safety limit will not be exceeded. TS 3.1.1 TS 3.1.1 Excess Reactivity Reactivity Specifications Specifications The excess reactivity reactivity shall not exceed 5.6% Ak/k. Liklk.
- Basis As shown shown in chapter 4 oftheof the SAR, this amount of excess reactivity reactivity will provide the capability capability to operate operate the reactor at full power with experiments in place. The primary limitation providing providing reactivity reactivity safety, however, is the shutdown margin margin requirement discussed in the next specification.
TS 3.1.2 TS 3.1.2 Shutdown MarginMargin Specifications Spec ifi cati ons The reactor shall not be operated operated unless the shutdown margin provided by control rods shall be greater than 0.2% Ak/k with: 0.2% Liklk 1.
- 1. the highest worth non-secured experiment in its most reactive state, non-secured experiment
- 2. the highest worth control element and the regulating blade (if not scrammable) fully withdrawn, and
- 3. the reactor reactor in the cold condition without xenon xenon..
- UWNR Safety Analysis Report Rev. 2 14-13 14-13 2008 Sept. 2008
Basis The value value of of the shutdown margin assures assures that the reactor can can be shut down from any operating condition even if the highest worth control element should regulating blade is not remain in the fully withdrawn position. If the regulating down
- scrammable, its worth is not used in determining the shutdown reactivity.
TS 3.1.3 TS 3.1.3 Pulse Limits Specifications Specifications
- 1. The reactivity to be inserted for pulse operation shall be determined and 1.
mechanically limited such that the reactivity insertion will not exceed
~klk.
1.4% Ak/k. 1.4%
- 2. Pulses
- 2. Pulses shall not be initiated at power levels exceeding exceeding 1 kilowatt kilowatt. .
Basis Measurements performed on the Puerto Rico Nuclear Center TRIGA-FLIP Measurements TRIGA-FLIP reactor indicated reactor that aa pulse insertion of reactivity indicated that reactivity of 1.4% 1.4% A k/k resulted in a
~ klk
- maximum temperature rise of approximately 400'C. With an ambient water approximately 400°C.
temperature of approximately temperature 100°C, the maximum approximately 100°C, maximum fuel temperature temperature would be approximately approximately 500'C500°C resulting in a safety margin margin of 500'C 500°C for standard fuel and 650°C 650'C for FLIP type fuel. Tests done on the thy mixed and all-FLIP cores 3,4,5,6 indicate that the fuel temperatures 3,4,5,6 indicate that the fuel temperatures measured measured and calculated calculated for the core arrangements arrangements allowed by these specifications do not exceed exceed 683°C in the worst case allowed. The temperature rise from pulse initiation is in addition addition to the temperature temperature in the fuel at the time the pulse is initiated. Limiting Limiting the initial initial power level level to 1I kW assures that excessive excessive temperatures temperatures will not be reached. These margins allow These margins amply for uncertainties allow amply uncertainties due to the accuracy accuracy ofof measurement measurement or location of the instrumented instrumented fuel element or due to the extrapolation extrapolation of data from the PRNC reactor. UWNR UWNR Safety Safety Analysis Report Rev. 2 14-14 14-14 Sept. 2008 2008
*
- TS TS 3.1.4 Core Configurations Configurations Applicability Applicability This specification applies applies to the configuration configuration of fuel and in-core experiments.
experiments. Objective Objective objective is to assure that provisions are made to restrict the arrangement The objective of fuel elements elements and experiments experiments so as to provide provide assurance that excessive power power densities will not be produced. Specifications Specifications
- 1. The core shall be an arrangement
- 1. arrangement of TRIGA uranium-zirconium uranium-zirconium hydride fuel-moderator bundles fuel-moderator bundles positioned in the reactor reactor grid plate.
- 2. The TRIGA core assembly may be standard, FLIP, or a combination, thereof (mixed core) provided that any FLIP fuel be comprised comprised of at least least thirty-five (35)
(35) fuel elements, located located in a contiguous, contiguous, central region.
- 3. The reactor shall not be operated with a core lattice position vacant except
- for positions on the periphery of the core assembly.
- 4. The reflector, excluding excluding experiments and experimental water or a combination of graphite and water.
experimental facilities, shall be
- 5. Fuel shall not be inserted or removed from the core unless the reactor is 5.
subcritical by more than the calculated calculated worth of the most reactive fuel assembly.
- 6. Control elements elements shall not be manually removed from the core unless the core has been shown to be subcritical with all control elements in the full out position.
position .
- UWNR Safety Safety Analysis Report Report Rev. 2 14-15 14-15 Sept. 2008 2008
Basis 1.
- 1. Standard TRIGA cores have been in use for years and their characteristics III all-FLIP cores have operated operated and their characteristics characteristics are well documented. The Puerto Rico Nuclear Center and the Gulf Mark characteristics are available.
- Gulf has also performed performed a series of experiments experiments using standard and FLIP fuel in mixed cores and a mixed core has been used successfully successfully in the Texas A&M University TRIGA reactor. In addition, studies performed performed at Wisconsin for a variety of mixed core arrangements arrangements indicate that such cores with mixed loadings would safely satisfy all operational requirements (SAR (SAR Chapters 4 and 6).
- 2. In mixed cores, it is necessary to arrange FLIP elements in a contiguous, contiguous, central region of the core to control flux peaking peaking and power generation generation peak values in individual elements.
- 3. Vacant core lattice positions will contain experiments or an experimental experimental facility to prevent prevent accidental fuel additions to the reactor core. They will be permitted only on the periphery periphery of the core to prevent power perturbations in regions of high power density.
perturbations
- 4. The core will be assembled assembled in the reactor grid plate which is located in a
- of light water. Water in combination pool oflight combination with graphite reflectors reflectors can be used for neutron economy and the enhancement enhancement of experimental experimental facility radiation radiation requirements.
5-6. Manual Manual manipulation of core components components will be allowed only when a single manipulation manipulation can not result in inadvertent inadvertent criticality. TS 3.1.5 TS 3.1.5 Reactivity Reactivity Coefficients Does not apply to TRIGA and TRIGA-FLIP TRIGA-FLIP reactors. UWNR Safety Analysis Report UWNR Report Rev. 2 14-16 14-16 2008 Sept. 2008
*
- TS 3.1.6 TS 3.1.6 Fuel Parameters Applicability Applicability This specification applies to the dimensional specification applies fuel elements.
dimensional and structural structural integrity integrity of the Objective Objective The objective is to assure that the reactor reactor will not be operated with defective fuel elements installed. Specifications Specifications The reactor reactor shall not be operated operated with damaged damaged fuel except for purposes of of identifying the damaged damaged fuel. A fuel element element shall considered damaged shall be considered damaged and must be removed removed from the core if: if:
- 1. In measuring sagitta77 exceeds 0.125 inch over the measuring the transverse bend, its sagitta length of the cladding;
- 2. In measuring the elongation, its length of the cladding exceeds exceeds its original length by 0.125 inch; and 3.
- 3. A clad defect exists as indicated indicated by detection of release release of fission products.
- 4. The fuel has not been visually inspected within the previous 15 months.
5.
- 5. The bumup burnup ofuranium-235 of uranium-235 in the UzrH fuel matrix matrix shall not exceed 50 exceed 50 percent 889 percent of the initial concentration.
concentration. ' ,9 Basis The limit of transverse bend has beenbeen shown shown to result in no difficulty in in disassembling the core. Analysis of the removal of heat from touching fuel elements shows that there will be no hot spots resulting in damage to the fuel caused by this touching. Experience Experience with TRIGA reactors reactors has shown that fuel element bowing that could result in touching has occurred without deleterious effects. The elongation limit has been specified specified to assure that the cladding cladding material will not be subjected to stresses stresses that could cause cause a loss of integrity in containment and to assure adequate the fuel containment adequate coolant flow throughthrough the top grid grid plate.. plate
- UWNR Safety Analysis Analysis Report Rev. 2 14-17 14-17 2008 Sept. 2008
TS 3.2 TS 3.2 Reactor Control and Safety Systems TS 3.2.1 TS 3.2.1 Operable Control Rods Operable Applicability Applicability
- This specification specification applies to the number operable control elements that number of operable must exist in order to operate the reactor.
Objective The objective objective of this requirement requirement is to insure that the reactor may be shut down from any condition condition of operation. Specifications Specifications The reactor shall not be operated operated unless at least three control elements are functioning and scrammable. scrammable. Basis
- In most cores the limits on shutdown shutdown margin actually dictate the number of of operable control elements required. Non-pulsing Non-pulsing cores do not require presence presence of a transient transient control rod if the shutdown shutdown margin requirements are met by the control blades.
UWNR Safety Analysis Report Rev. 2 14-18 14-18 Sept. 2008 2008
*
- TS 3.2.2 TS 3.2.2 Reactivity Insertion Rates Applicability Applicability This specification Rates (Scram time) specification applies to the time required for the scrammable control elements elements to be fully inserted from the instant that a safety channel variable reaches reaches the Safety Safety System Setting.
Objective The objective is to achieve prompt shutdown of the reactor to prevent fuel damage. Specifications The scram time measured from the instant a simulated signal reaches the value of the LSSS to the instant that the slowest scrammable scrammable control element reaches its fully inserted inserted position shall not exceed 2 seconds. Basis
- This specification specification assures that the reactor will be promptly promptly shut down when a scram signal is initiated. Experience Experience and analysis analysis have indicated that for the range of transients anticipated anticipated for a TRIGA reactor, the specified scram time is adequate to assure the safety of the reactor.
TS 3.2.3 TS 3.2.3 Other Pulsed Operation Limitations Limitations other than those on core configuration configuration and pulsed reactivity insertion insertion limits are not required required on this reactor.
- UWNR Safety Analysis Report Rev. 2 14-19 14-19 2008 Sept. 2008
TS TS 3.2.4 Reactor Safety System Applicability Applicability System This specification applies to the reactor reactor safety system channels.
- Objective Objective The objective is to specify the minimum number of reactor safety channels channels that must be operable for safe operation.
Specifications Specifications The reactor shall not be operated operated unless the safety channels described in Table Table 3.2.4 are operable. Table Ta bl e 3.2.4 Reactor Reactor Safety Safety System Channels Channels Number Number operable operable Safety Channel Setpoint Setpoint and Function in specified mode SS SW PU PU Fuel Temperature Scram if fuel temperature temperature exceeds >400'C the fuel temperature temperature safety
>400°C in channel.
event of loss of all available fuel thermocouples thermocouples and inability to obtain a replacement replacement instrumented fuel element, In the 1 1 1
- operation operation may continue in any operational core if the linear power level scram points are reduced to are reduced I110% full to 110% full power.
power. Linear Power Level Scram if power > 125%125% full power 2 22 -- Manual Scram Manually Manually initiated scram 1 1 1 Preset Timer Transient rod scram 15 seconds or less after - - 1 pulse Reactor water level Scram if < 19 feet above top of core 1 1 1 High Voltage Monitor Scram on loss of high voltage to neutron and Scram 1 1 1 instrument detectors gamma ray power level instrument UWNR UWNR Safety Analysis Analysis Report Rev. 2 14-20 2008 Sept. 2008
- 0
Basis
- The fuel temperature and power scrams protection to ensure that the scrams provide protection before the safety limit on fuel temperature is reached.
reactor is shut down before required because FLIP fuel is no longer manufactured. If a The exception is required core has been tested to meet the definition operational core the power definition of an operational power level scrams provide adequate adequate protection protection to assure the LCO of fuel temperature is not exceeded. temperature operator a means of rapid shutdown in the event The manual scram allows the operator of unsafe or abnormal conditions. The preset timer assures reduction of reactor power to a low level after a pulse. The reactor shutdown of the reactor in the reactor pool water level scram assures shutdown event of a serious leak in the primary system or pool. reactor with other systems The high voltage monitor prevents operation of the reactor inoperable due to failure of the detector high voltage supplies inoperable supplies..
- Analysis Report Rev. 2 UWNR Safety Analysis 14-21 2008 Sept. 2008
TS TS 3.2.5 3.2.5 Interlocks Applicability Applicability This section applies to the interlocks interlocks which inhibit or prevent withdrawal or reactor startup. prevent control element
- Objective Objective The objective of these interlocks is to prevent operation under unanalyzed objective ofthese unanalyzed or imprudent conditions.
Specifications Specifications The reactor reactor shall not be operated in the indicated modes unless the the. interlocks interlocks in Table 3.2.5 are operable. Table 3.2.5 Interlocks Interlocks Number operable operable Channel Setpoint and Function in specified mode SS SW PU Log Count Rate Transient Rod Control Prevent control element withdrawal withdrawal when neutron count rate < 2 per second Prevent application application of air to fire transient rod unless drive is at IN limit. 1 1 1 0 1 0
- Log N Power Level Prevent application application of air to fire transient rod 1 1 1 power level is above 1I kW and when power transient rod is not full in.
Pulse Mode Control Prevents withdrawal of control blades while 0 0 1 in pulse mode. Basis The Log count rate interlock does not allow control element withdrawal unless the neutron neutron count rate is high enough to assure proper instrument' response response during reactor startup. The Transient Rod Control interlock prevents inadvertent addition of of excessive amounts or reactivity reactivity in steady-state steady-state modes. The Log N interlock prevents firing of the transient rod at power levels above
- 1.0 kW if the transient transient rod drive is not in the full down position. This UWNR Safety Analysis Report Rev. 2 14-22 14-22 Sept. 2008 2008
effectively prevents effectively inadvertent pulses which might cause fuel temperature prevents inadvertent temperature to
- exceed the safety limit on fuel temperature temperature..
The pulse mode control blade withdrawal interlock interlock prevents prevents reactivity addition in pulse mode other than by firing the transient rod. TS 3.2.6 TS 3.2.6 Backup Shutdown Mechanisms Backup shutdown mechanisms mechanisms are not required required for this reactor. TS 3.2.7 TS 3.2.7 Bypassing Channels Applicability Applicability This specification specification applies to the interlocks interlocks in Table 3.2.5. 3.2.5. Objective Objective The objective objective is to indicate the conditions conditions in which an interlock may be be bypassed.
- Specifications Specifications The Log Count Rate interlock 1.
interlock in Table 3.2.5 may be bypassed:
- 1. During fuel loading in order to allow control element withdrawal necessary necessary for the fuel loading procedure procedure or or
- 2. When Log Power Level and Linear Power Level Level channels channels are on-scale.
Basis During early stages of fuel loading the count-rate count-rate on the source range channel will be below the interlock setpoint. The bypass allows control control element movements necessary movements necessary for loading fuel with control elements partially partially withdrawn withdrawn and for performing inverse multiplication multiplication determinations determinations of control element element worth and core reactivity status. Once the other power indications indications are available the startup count rate channel channel is no longer required, so the interlock interlock no longer serves any purpose. purpose .
- UWNR UWNR Safety Analysis Report Rev. 2 14-23 Sept. 2008 2008
TS 3.2.8 TS 3.2.8 Control Systems Control Systems and Instrumentation Instrumentation Required for OperationOperation Applicability Applicability This specification reactor specification applies to the information which must be available reactor operator during during reactor reactor operation. available to the
- Objective The objective objective is to require require that sufficient sufficient information is available to the operator to assure safe operation ofthe of the reactor.
Specifications Specifications The reactor reactor shall not be operated unless measuring channelschannels listed in Table 3.2.8 are operable. Table .. IInstrumentation 3.2.8 T able 328 nstrumentaf IOn andandC ontro I s R Controls . d £for Required eqUlre 0'peraf IOn or Operation Number operable Number operable Channel Channel Function in specified specified mode SS SW PU PU Fuel Temperature Temperature
,
Input for fuel temperature temperature scram. In the event of loss of all available fuel thermocouples and inability to obtain a replacement operation replacement operational operation may continue in any operational core. 1 1 1
- 0 Linear Power Level Input for safety system power level scram j22 2 0 Log Power Level Wide range power power indication, indication, permissive permissive for 1 1 0 initiation of Pulse Mode Startup Log Count Rate Wide range power power indication, permissive permissive for 1*
1* 1* 1* 0 control element element withdrawal Pulsing Power Level Pulsing Level Pulse power level indication indication 0 0 1 Required during startup only until the Log Power Level and Linear Power Level channels
- Required during startup only until the Log Power Level and Linear Power Level channels are
- are on-scale Basis Fuel temperature temperature indicated at the control console gives continuous information continuous information on the process variable variable which has a specified specified safety limit. The exception exception isis required required because FLIP fuel is no longer manufactured. If a core has been tested to meet the definition of an operational core the power level scrams UWNR Safety Analysis Report Rev. 2 14-24 14-24 2008 Sept. 2008
- provide adequate protection to assure the LCO of fuel temperature temperature is not
- exceeded exceeded..
reactor power level The power level monitors assure that reactor level is adequately monitored for all modes of operation. monitored TS 3.3 Reactor Pool Water Systems TS 3.3 Systems Applicability This specification specification applies to the pool containing the reactor and to the cooling cooling of the core by the pool water. Objective Objective The objective is to assure that coolant water shall be available to provide adequate cooling of the reactor core and adequate radiation shielding and to prevent damage damage to in-pool components components by corrosion. Specifications
- l. reactor core shall be cooled by natural
- 1. The reactor natural convective water flow flow..
- 2. The pool water inlet pipe to the demineralizer demineralizer demineralizer shall demineralizer shall be equipped with a check inadvertent inadvertent draining of the pool.
shall not extend more than 15 feet into the top of the reactor pool when fuel is in the core. The outlet pipe from the check valve and siphon breaker to prevent prevent
- 3. other auxiliary systems pumps shall be located no more than 15
- 3. Diffuser and other 15 feet below the top of the reactor pool.
entering the pool shall have siphon
- 4. All other piping and pneumatic tube systems entering siphon breakers and valves or blind flanges which will prevent draining more than 15 feet breakers of water from the pool.
coolant if the pool level drops one foot or
- 5. A pool level alarm shall indicate loss of coolant 5.
less below normal level. operated if the conductivity
- 6. The reactor shall not be operated conductivity of the pool water water exceeds exceeds 5 micromhos/cm (<0.2 micrornhos/cm <<0.2 MegOhm-cm) when averaged over a period of one week.
- 7. The reactor shall not be operated radioactivity of pool water exceeds operated if the radioactivity exceeds the Appendix B Table 33 for radioisotopes with half-lives limits of 10 CFR Part 20 Appendix half-lives
- hours..
>24 hours UWNR UWNR Safety Analysis Analysis Report Report Rev. 2 14-25 Sept. 2008 2008
Basis
- 1. This specification specification is based on thermal and hydraulic calculations which show that the TRIGA-FLIP TRIGA-FLIP core can operate operate in a safe manner at power levels up to 2,706 2,700 0
- kW with natural convection convection flow of the coolant water. A comparison comparison of operation operation of the TRIGA-FLIP TRIGA-FLIP and standard TRIGA TRIGA Mark Mark III has shown operation operation to be safe safe for the above power power level. Thermal Thermal and hydraulic characteristics of mixed cores hydraulic characteristics are essentially the same asthat as that for TRIGA-FLIP TRIGA-FLIP and standard corescores...
- 2. The inlet pipe to the demineralizer demineralizer is positioned so that a siphon action will drain less than 15 feet of water. The outlet pipe from the demineralizer demineralizer discharges discharges into a pipe entering the bottom of the pool through a check valve which prevents leakage from the pool by reverse flow from pipe ruptures or improper operation operation of of the demineralizer demineralizer valve manifold. In addition, the pipe has a loop equipped with a siphon breaker breaker which which prevents prevents loss of pool water.
- 3. In the event of pipe failure and siphoning siphoning of pool water, the pool water water level will drop no more than 15 feet from the top of the pool.
- 4. Other pipes which enter the pool have siphon breakers which prevent pool drainage. Valves are provided provided for pneumatic tube system lines and primary cooling system pipe. Other Other piping installed in the pool has blind flanges flanges
- permanently installed.
permanently
- 5. Loss of coolant alarm, after one foot of loss, requires corrective
- 5. corrective action. This alarm is observed observed in the reactor reactor control room and outside the reactor building.
- 6. The conductivity limit assures that materials materials within within the pool will not be degraded degraded and that the radioactivity radioactivity of the pool water will be minimized.
- 7. Analyses in section section 12.2.9 of the Safety Analysis Report show that limiting the Analysis Report activity to this level level will not result in any person being exposed to concentrations conceritrations greater greater than those permitted pennitted by 10 CFR Part Part 20.
TS 3.4 Confinement TS Confinement Applicability Applicability These specifications specifications apply to the room housing the reactor and the ventilation system ventilation system controlling controlling that room. Objective Objective The objective is to provide restrictions on release of airborne radioactive materials to B
- the environs.
UWNR Safety UWNR Safety Analysis Analysis Report Report Rev. 2 14-26 2008 - Sept. 2008
Specifications
- 1. The reactor shall be housed in a closed room designed to restrict leakage. The minimum minimum free volume shall be 2,000 cubic meters.
- 2. All air or other gas exhausted exhausted from the reactor room and associated facilities shall be released to the environment associated experimental environment a minimum of 26.5 meters meters above ground level.
Basis Calculations in Chapter 13 of the Safety Calculations Safety Analysis Report show that exposure of of occupants of the Laboratory can be kept below occupants below 10 CFR Part 20 limits for occupational occupational exposure under accident conditions if the room volume is 2,000 m exposure m33 .* Calculations Calculations in Chapter 13 13 of the SAR based on release of radioactive radioactive effluent at ground ground level show show that concentrations concentrations of radioactive materials materials are within limits of 10 CFR Part 20 for non-restricted areas during the accidents accidents considered. Further calculations calculations based onon release at the stack height show a further reduction reduction by a factor of 2.6 due to operation of2.6 operation of the ventilation system and release of effluent at a height of26.5m. of 26.5m. TS 3.5 TS 3.5 Ventilation Systems Systems
- Applicability This specification specification applies to the operation of the reactor Objective reactor laboratory laboratory ventilation system.
The objective is to assure that the ventilation ventilation system is in operation to mitigate mitigate the consequences of the possible release of radioactive materials resulting from reactor consequences reactor operation. Specifications Specifications The reactor shall not be operated unless the laboratory laboratory ventilation system is in operation, operation, except for periods of time not to exceed two days, to permit repairs of the system. Basis It is shown in the SAR Chapter 11 that Argon-41 release at zero stack height results in in concentrations con~entrations less than the concentrations concentrations permitted for non-restricted non-restricted areas. calculations indicate that operation Further, the calculations operation of the ventilation ventilation system reduces the
- concentration to which the public would be exposed by a factor of 10 below this limit concentration limit..
Exposures in the event of a fuel element cladding leak are also calculated based on on UWNR UWNR Safety Analysis Report Rev. 2 14-27 14-27 2008 Sept. 2008
non-operation of the ventilation system. Therefore, Therefore, operation of the reactor with the TS ventilation system shut down in order to make repairs assures the degree of control on calculations are, which the calculations TS 3.6 Emergency Emergency Power Power are. based. on
- Emergency power systems are not required Emergency required for this facility.
3.7 Radiation TS 3.7 Radiation Monitoring Monitoring Systems and Effluents TS 3.7.1 Monitoring Systems Monitoring*Systems Applicability Applicability This specification specification applies to the radiation monitoring information which must be available available to the reactor operator operator during reactor reactor operation. Objective The objective is to assure that sufficient radiation monitoring information information is
- available available to the operator to assure safe operation operation of the reactor.
Specifications Specifications The reactor shall not be operated operated unless the radiation monitoring monitoring channels channels listed in Table 3.7.1 are operable. operable. Table .. 1 Radiation Tabl e 3.7.1 37 R ad*laf IOn M om*t* onng S Monitoring )yst ems Systems Radiation Monitoring Function Number Channels** Channels Area Radiation Monitor Monitor radiation radiation levels within the reactor reactor room 3 Exhaust Gas Radiation Monitor radiation radiation levels in the exhaust air stack stack 1 Monitor Monitor , Exhaust Particulate Particulate Radiation Monitor Monitor radiation levels in the exhaust exhaust air stack stack 11 Monitor Monitor Environmental Environmental Radiation TLD dosimeters evaluated on a quarterly basis 4 Monitors Monitors record exposure surrounding the stack exposure in area surrounding stack
- For periods of time for maintenance to the radiation monitoring channels, the intent of this
- For periods of time for maintenance to the radiation monitoring channels, the intent of this specification will be satisfied satisfied if they are replaced with portable gamma sensitive instruments having their own alarms or which shall be kept under visual observation.
UWNRSafety Analysis Report Rev. 2 UWNRBafety 14-28 2008 Sept. 2008
- Basis
- The radiation monitors provide information impending of radioactivity radioactivity to the surroundings.
in areas immediately information to operating personnel of any impending or existing danger from radiation so that there will be sufficient time to evacuate the facility and take the necessary necessary steps to prevent surroundings. The environmental sufficient prevent the spread environmental monitors are placed surrounding the reactor laboratory immediately surrounding laboratory to record actual dose that would have beenbeen delivered delivered to a person continually present in the area. TS 3.7.2 TS Effluent (Argon-41) Effluent (Argon-41) Discharge Limit Limit Applicability This specification specification applies to the concentration concentration of Ar-41 which may be discharged from the facility. Objective The objective is to insure that the health and safety of the public are not ofthe endangered by the discharge of Ar-41. endangered Ar-41. Specifications
- The concentration concentration of Ar-41 in the effluent gas from the facility, as diluted by atmospheric atmospheric air in the lee of the facility as a result of the turbulent wake effect, shall not exceed 1lxi 0'88 !-lCi/ml x 10- [Ci/ml averaged averaged over one year.
Basis 10 CFR Part 20 Appendix Appendix B, Table II I1specifies a limit of 1 x 10-10.'8 !-lCi/ml ptCi/ml for Ar-41. Ar-41. Chapter Chapter 11 and Appendix Appendix A of the SAR substantiates substantiates a release level ofof 3.3E-9 p.Ci/ml
!-lCi/ml for a 3.54 meters/second (lowest (lowest monthly monthly average) wind speed if all Ar-41 produced were continuously continuously discharged. The dilution factor by which emitted material is diluted is 2.5E-4 !-lCi/ml
[tCi/ml per Ci/second discharged.
- UWNR Safety Analysis Report Rev. 2 14-29 Sept. 2008
TS 3.8 TS 3.8 Experiments Applicability This specification specification applies to experiments installed in the reactor and its experimental facilities.
- W Objective The objective is to prevent damage to the reactor or excessive release of radioactive materials in the event of an experiment failure.
TS 3.8.1 TS 3.8.1 Reactivity Reactivity Limits Specifications Specifications The reactor shall not be operated unless the following conditions governing experiments exist:
- 1. The reactivity
- 1. reactivity worth of any single non-secured non-secured experiment shall not k/k.
A kJk. exceed 0.7% !:::.
- 2. The reactivity worth of any single secured experiment 1.4% A k/k.
1.4%!:::. kJk. Basis experiment shall not exceed
*
- 1. This specification
- 1. specification is intended intended to provide assurance that the worth of a single unfastened experiment will be limited to a value such that the safety unfastened experiment limit will not be exceeded exceeded if the positive worth of the experiment experiment were to be suddenly suddenly inserted (SAR (SAR Chapter Chapter 13).
13).
- 2. The maximum maximum worth of a single experiment is limited so that its removal single experiment from the cold critical reactor from the cold critical reactor will not result in the reactor achieving achieving a power level high enough enough to exceed exceed the core temperature temperature safety limit.
Since experiments of Since experiments such worth must be fastened in place, its removal from the reactor operating operating at full power wouldwould result in a relatively relatively slow slow power increase increase such that the reactor reactor protective protective systems systems would act to prevent high power levels from being attained. prevent high power levels from being attained. SAR accident analysis analysis includes includes a sudden addition 1.4% A k/k addition of 1.4%!:::. k/k from firing the transient transient control rod while while operating operating at thethe power power level level scram scram point, a more more severe severe transient transient result from removal of aa: fixed experiment than that which could result experiment with with the 0 same reactivity worth. UWNR Safety Analysis UWNRSafety Analysis Report Rev. 2 14-30 14-30 Sept. 2008 2008
- TS 3.8.2 TS 3.8.2 Materials
- Specifications
- 1. Explosive materials,
- 1. materials, such as gunpowder, TNT, nitroglycerin, or PETN, in quantities quantities greater than 25 milligrams milligrams shall not be irradiated in the reactor or experimental experimental facilities. Explosive materials in quantities quantities less than 25 milligrams may be irradiated provided provided the pressure produced produced upon detonation of the explosive has been calculated calculated and/or experimentally experimentally demonstrated to be less than the design pressure of the container.
demonstrated
- 2. Experiment materials, except fuel materials, materials, which could off-gas, sublime, volatilize, or produce produce aerosols under (1)
(1) normal operating conditions conditions ofof experiment or reactor, (2) the experiment (2) credible accident accident conditions in the reactor, or (3) possible accident accident conditions in the experiment experiment shall be limited in activity such that if 100% 100% of the gaseous activity or radioactive aerosols aerosols produced escaped to the reactor room or the atmosphere, atmosphere, the airborne concentration of radioactivity radioactivity averaged over a year would not exceed the limit of Appendix Appendix B of 10 CFR Part 20.
- 3. In calculations calculations pursuant to 2 above, the following assumptions shall be used:
- a. If the effluent effluent from an experimental tank which which closes experimental facility exhausts through a holdup automatically on high radiation level, closes automatically the gaseous activity or aerosols produced will escape.
- b. If the effluent effluent from an experimental 10% of level, at least 10%
experimental facility exhausts through a filter of installation installation designed for greater than 99% efficiency for 0.3 micron 99% efficiency micron particles, at least 10% 10% of these vapors can escape.
- c. For materials whose boiling point is above 130°F 1307F and where vapors formed by boiling this material can escape escape only through an undisturbed undisturbed column column of water above the core, at least 10%10% of these vapors can escape.
escape.
- d. An atmospheric atmospheric dilution factor of2.5 10'4 !lCi/ml of 2.5 x 10- ýiCi/ml per Ci/s for gaseous discharges discharges from the facility.
- 4. Each fueled experiment experiment shall be controlled controlled such that the total inventory inventory ofof iodine isotopes 131 through 135 in the experiment experiment is no greater than 1.51.5 curies.
cunes .
- UWNR Safety Analysis Report Rev. 2 14-31 14-31 Sept. 2008
Basis
- 1. This specification 1.
2-3. 2-3. specification is intended to prevent resulting from failure of an'experiment These specifications damage to reactor components prevent damage an experiment involving explosive materials. specifications are intended to reduce reduce the likelihood that airborne
- activities in excess of the limits of Appendix B of 10 CFR Part 20 will be released to the atmosphere outside the facility boundary boundary of the UWNR. The dilution factor is based on computations computations reported in Chapter 11 11 and Appendix Appendix A of the Safety Safety Analysis Report.
- 4. The 1.5 curie limitation on iodine 131 through 135 assures that in the of a fueled experiment event of failure ofa experiment leading to total release of the iodine, the exposure dose at the exclusion area boundary boundary will be less than
. that allowed by 10 CFR Part 20 for an unrestricted unrestricted area.
TS TS 3.8.3 3.8.3 Experiment Failure and Malfunctions Experiment Malfunctions Specifications Specifications If a capsule fails and releases releases material material which could damage damage the reactor fuel or structure structure by corrosion corrosion or other means, removal and physical inspection of the capsule shall be performed performed to determine the consequences consequences and need for
- corrective corrective action. The results of the inspection and any corrective action taken shall be reviewed reviewed by the Reactor Director or his designated designated alternate and determined determined to be satisfactory satisfactory before before operation of the reactor reactor is resumed.
Basis Operation Operation of the reactor with a failed capsule is prohibited prohibited to prevent prevent damage to the reactor fuel or structure. Failure of a capsule must be investigated to assure no damage has or will occur. TS 3.9 Facility Specific TS 3.9 Specific LCOs LeOs There are no facility specific specific LCOs at this facility. UWNR Safety UWNR Safety Analysis Analysis Report Report Rev. 2 14-32 14-32 2008 Sept. 2008
- TS TS44 SURVEILLANCE REQUIREMENTS SURVEILLANCE REQUIREMENTS
- In accordance accordance with section 4.0 of Standard ANSI/ANS-15.1, average surveillance times between surveillance the specification.
ANSIIANS-15.1, the following terms for surveillance intervals shall allow, for operational operational flexibility only, maximum surveillance intervals as indicated below unless otherwise specified within
** Five-year interval not to exceed six years.
** Biennial interval not to exceed exceed two and one-half years.
- Annual interval not to exceed 15 months.
** Semiannual interval not to exceed exceed seven and one-half months.
- Quarterly Quarterly interval not to exceed exceed four months.
** Monthly interval not to exceed six weeks.
- Weekly interval not to exceed ten days
- Daily interval must be done within the calendar day.
Scheduled surveillances, Scheduled surveillances, except those specifically specifically required required when the reactor is shut down, may be deferred during shutdown shutdown periods, but be completed prior to subsequent reactor reactor startup unless operation is required for the performance performance of the surveillance. Scheduled Scheduled operating may be deferred until surveillances which cannot be performed with the reactor operating planned reactor shutdown. If the reactor is not operational a planned operational in a particular mode, surveillances required specifically for that mode may be deferred surveillances deferred until the reactor becomes operational operational in that mode. mode .
- General Applicability Applicability This specification specification applies to the surveillance safety.
surveillance requirements of any system related to reactor reactor Objective Objective The objective is to verify verify the proper operation operation of any system related to reactor safety after maintenance maintenance or modification modification of the system. Specifications Any additions, modifications, or maintenance maintenance to the ventilation ventilation system, the core and its associated support structure, the pool or its penetrations, penetrations, the pool coolant system, the rod drive mechanism, or the reactor safety system shall be made and tested in accordanceaccordance with the specifications specifications to which the systems systems were originally designed and fabricated fabricated or to specifications specifications approved by the Reactor Safety Safety Committee. A system shall not be be considered operable considered operable until after it is successfully successfully tested tested..
- UWNR UWNR Safety Analysis Report Rev. 2 14-33 14-33 ,I Sept. 2008
Basis This specification relates to changes changes in reactor systems which could directly directly affect the safety of the reactor. As long as changes or replacements to these systems continue to meet the original design specifications, specifications, then it can be assumed that they meet the presently accepted operating criteria.
- TS TS 4.1 Reactor Reactor Core Parameters Applicability Applicability These specifications specifications apply to the surveillance requirements for measurements, tests, calibrations of reactor core parameters.
and calibrations parameters. Objective The objective is to verify the core parameters parameters which are directly related to reactor reactor safety. Specifications Specifications
*
- 1. Excess reactivity 1.
reactivity shall be determined Excess reactivity determined at least annually and after changes changes in either the core, in-core experiments, experiments, or control elements for which the predicted change in reactivity exceeds the absolute value of the specified shutdown margin.
- 2. Shutdown margin The shutdown shutdown margin margin shall be determined determined at least annually and after changes changes in either the core, in-core experiments, experiments, or control elements.
- 3. Pulse limits The reactor shall be pulsed semiannually to compare compare fuel temperature measurements measurements (if an operating fuel thermocouple thermocouple is available) and peak power levels with those of previous pulses of the same reactivity reactivity value.
- 4. Core configuration configuration Each planned change in core configuration configuration shall be determined determined to meet the requirements of Sections 3.1 3.1(4)
(4) and 5.3 of these specifications specifications before the core is loaded.
- 5. Reactivity Coefficients
- 5. Coefficients Power defect and pulsing characteristics characteristics shall be measured measured during startup UWNR testing of cores containing predictions UWNR Safety Analysis Report Rev. 2 containing different different fuel compositions predictions in the Safety Analysis Report.
14-34 14-34 compositions and compared compared to
'
Sept. 2008 2008
*
- 6. Fuel Parameters
- a. All fuel elements shall be inspected visually for damage or deterioration annually.
deterioration Uninstrumented fuel elements which have been resident in the core during the
- b. Uninstrumented sagitta annually. Fuel elements previous year shall be measured for length and sagitta shall not be added sagitta has added to a core unless a measurement of length and sagitta been completed within previous fifteen months.
within the previous
- c. Fuel elements in the hottest assumed location, as well as representative representative elements in each of the rows, shall be measured for possible damage in the indication that the Limiting Safety System Setting may have event there is indication been exceeded.
Basis 1-2. Annual measurements, measurements made after changes that measurements, coupled with measurements can affect affect reactivity values provide adequate assurance that core behavior will adequate assurance be as analyzed. The reactivity values in FLIP fuel change change very slowly with burnup. fuel bumup. changes in behavior are resulting from fuel verifications assure no changes
- 3. Semiannual verifications
- characteristic inadvertent by composition changes..
characteristic changes
- 4. Checking contemplated contemplated core configurations composition restrictions.
requirements will prevent configurations against requirements inadvertent loading of cores which do not meet power peaking peaking restraints imposed restraints imposed Measurements made during core startup testing are sufficient to assure core
- 5. Measurements 5.
behavior will be as analyzed. behavior inspection of the TRIGA fuel has been shown
- 6. Annual inspection shown adequate adequate to assure fuel element integrity through a long history of standard operation operation..
- UWNR Safety Analysis Report Rev. 2 UWNR 14-35 14-35 Sept. 2008 2008
TS TS 4.2 Reactor Control and Safety Systems Applicability specifications apply to the surveillance These specifications and calibrations of surveillance requirements for measurements, tests, calibrations the control and safety systems.
- Objective Objectiv,e The objective is to verify the performance performance and operability of those systems and components components which are directly related to reactor safety.
Specifications 1.
- 1. Reactivity Reactivity worth of\ control control elements elements The reactivity worth of control elements shall be determined determined upon substantiative composition or arrangement substantiative changes in core composition arrangement and annually thereafter.
- 2. Control element withdrawal withdrawal and insertion speeds Control element drive withdrawal withdrawal and insertion insertion speeds shall be measured annually and following maintenance maintenance to the control element or the control control
- element element drive mechanism.
- 3. Transient Rod and Associated Mechanism Associated Mechanism The transient rod drive cylinder cylinder and associated air supply system shall be inspected, cleaned, and lubricated necessary annually.
lubricated as necessary
- 4. Scram times of control and safety elements The scram time for all scrammable scrammable control control elements elements shall be measured measured annually and following maintenance maintenance to the control elements or their drives.
- 5. Scram Scram and Power Measuring Channels
- a. A channel test of each Reactor Reactor Safety System measuring measuring channel in Table (1) through (4) 3.2.4 items (1) (4) and the interlocks interlocks in Table 3.2.5 required required for the intended intended modes of operation shall be performed within 24 hours before each each day's operation or prior to each operation extending more than one day.
- b. A channel test of items (5) and,(6) in Table performed semi-Table 3.2.4 shall be performed annually.
- 6. Operability Tests This concern is covered by the General Surveillance This Surveillance criterion at the beginning UWNR UWNR Safety of this section.
Safety Analysis Analysis Report Report Rev. 2 14-36 14-36 2008 Sept. 2008
- 7.
- 7. Thermal Calibration-Forced Convection Thermal Power Calibration-Forced Convection
- Not applicable applicable to this reactor reactor
- 8. Thermal Power
- 8. Calibration-Natural Convection Power Calibration-Natural Convection A
A Channel Calibration Calibration shall shall be made of the power power level level monitoring monitoring channels channels by by the calorimetric method calorimetric method upon substantiative substantiative changes changes in core composition core composition arrangement and annually thereafter. or arrangement thereafter.
- 9. Control Element Inspection
- 9. Inspection The control elements elements shall be visually inspected for deterioration deterioration biennially.
biennially. Basis
- 1. Control
- 1. element worths change slowly Control element slowly unless the core arrangement is changed, core arrangement changed, so annual measurement measurement is sufficient sufficient to assure safety.
- 2. Control element element insertion insertion or withdrawal withdrawal speeds are fixed by by the motor design and and thus do not change change except for extreme binding binding conditions conditions within thethe drive.
drive.
- 3. Transient rod drive and air supply
- 3. and 'lubrication, so an annual supply includes filtration and,lubrication, coupled with pre-startup check coupled pre-startup checks checks is sufficient sufficient to assure assure operabilty.
operabilty.
- 4. Measurement Measurement of the scram time on an annual basis is a check check not only of the
- 5.
scram system electronics, rods to perform be checked checked electronics, perform properly.
- 5. The items 11 through properly.
but also is an indication indication of the through 4 in the table are essential safety equipment frequently, even though capability of capability of the the control equipment and thus should though no failures have been observed by should by checkout checkout in in nearly 5050 years years of operation. operation. Frequent unnecessary for item 5, Frequent testing is unnecessary 5, a simple float switch which is very very unlikely nearly 50 performed for nearly unlikely to fail, and has performed 50 years without a failure. Testing Testing item 6, 6, the high voltage voltage monitor scram, results in changing the voltage changing voltage to the neutron detectors. This introduces neutron detectors. introduces step changes into step changes the signal circuits of the measuring channels which can measuring channels can lead to long recovery times and a significant significant increase increase in failures of the measuring measuring channels. Further, channels. Further, since since the checkout checkout of the linear linear safety channels is a source channels source check, if high voltage voltage were lost that check check would not be possible possible if the voltage had been lost. 6.
- 6. The The general general requirement checks of equipment requirement for checks operability after equipment operability after maintenance maintenance or or modification modification of systems will reveal any loss of safety functions due to the maintenance maintenance or modification.
modification.
- 8. The power
- 8. calibration will assure that the reactor power level channel calibration reactor will be operated operated at the proper proper power levels.
- 9. Annual checks
- 9. checks in other reactors and for nearly 50 other TRIGA reactors 50 years in this reactor reactor have been sufficient have been sufficient to insure no failures due to deterioration.
deterioration. UWNR Safety Analysis Report Rev. 2 1-7Sp.20 14-37 14-37 2008 Sept. 2008
I TS Coolant Systems TS 4.3 Coolant Applicability Applicability This specification specification applies to the reactor reactor pool water.
- 0 Objective The objective is to assure the water quality and radioactivity radioactivity is within the defined defined limits Specifications Specifications The pool water conductivity and radioactivity radioactivity shall be measured measured quarterly.
Basis Pool water conductivity conductivity is continuously monitored, but would be manually monitored on a quarterly quarterly basis if the instruments instruments failed. Radioactivity Radioactivity is indirectly monitored monitored by an area radiation monitor near the demineralizer demineralizer bed, so gross activity increases would be detected immediately. Experience Experience with TRIGA TRIGA reactors reactors indicates indicates the earliest earliest detection of fuel clad leaks leaks is usually from airborne activity, rather rather than pool
- water activity. The quarterly measurement water(activity. measurement can identify specific specific radionuclides.
radionuclides. 0 TS TS 4.4 Confinement Confinement No surveillances surveillances are required. TS 4.5 Ventilation Systems TS Applicability Applicability This specification specification applies to the building building confinement ventilation system. Objective The objective is to assure the proper operation of the ventilation proper operation ventilation system in controlling releases of radioactive radioactive material material to the uncontrolled uncontrolled environment. Specifications Specifications It shall be verified quarterly and following repair or maintenance maintenance that the ventilation ventilation
- system is operable.
UWNR UWNR Safety Analysis Analysis Report Rev. 2 14-38 14-38 2008 Sept. 2008
Basis
- Over 30 years of experience with the previous that testing the system quarterly is sufficient system and control of the release expected to exceed is expected appropriate.
demonstrated previous ventilation system has demonstrated sufficient to assure the proper operation operati'on of the release of radioactive material. The new ventilation exceed the reliability of ventilation system system the previous system so quarterly testing is still ofthe TS 4.6 Emergency Electrical TS Electrical Power Systems Not Applicable. Applicable. TS TS 4.7 Radiation Monitoring SystemsSystems and Effluents TS 4.7.1 TS Radiation Monitoring Systems Radiation Systems Applicability specification applies to the surveillance requirements for the area This specification radiation monitoring equipment equipment and the stack air monitoring monitoring system. Objective
- The objective objective is to assure that the radiation monitoring and to verify verify the appropriate Specifications Specifications appropriate alarm alarm settings.
monitoring equipment equipment is operating The radiation monitoring and stack monitoring systems shall be calibrated calibrated annually and shall be verified to be operable by monthly source checkschecks or channel tests tests. . Basis Experience has shown that monthly verification of area radiation monitor Experience monitor operability operability and setpoints setpoints in conjunction conjunction with the downscale-failure downscale-failure feature ofof the instrument is adequate to assure operability. Annual calibration is adequate adequate to correct correct for any variation in the system due to a change of of operating characteristics over a long time span. Annual calibrations operating characteristics calibrations and monthly source source or channel channel checks checks of the stack particulate and gaseous monitors, along with the high or low flow alarms associated with the monitor assure operability and accuracy. accuracy .
- UWNR Safety Analysis Report Rev. 2 UWNR 14-3.9 14-39 2008 Sept. 2008
TS 4.7.2 TS Effluents Applicability Applicability specification applies to gaseous and liquid discharges This specification laboratory. discharges from the reactor reactor
- Objective Objective The objective objective is to assure that ALARA ALARA and 10 CFR Part 20 limits are observed.
Specifications Specifications Liquid radioactive radioactive waste discharged to the sewer system shall be sampled sampled for to assure levels are below applicable limits before discharge. radioactivity to* Results of the measurements shall be recorded recorded and reported in the Annual Report. The total annual release of gaseous radioactivity to the environment environment shall be recorded and reported in the Annual Report.
- Basis Liquid waste releases are batch releases, so the liquid can be sampled before release. Air activity activity discharged is continuously continuously recorded and the integrated integrated release is reported.
;
TS TS 4.8 Experiments No surveillances surveillances are required. TS TS 4.9 Facility-Specific Facility-Specific Surveillance Surveillance Not applicable. There is no facility-specific facility-specific surveillance. surveillance. ( UWNR Safety Analysis Report Rev. 2 14-40 14-40 Sept. 2008
- TS 55 TS DESIGN DESIGN FEATURES FEATURES
- TS 5.1 Site and Facility Description TS 5.1 Specifications Description
- 1. The reactor shall be housed in a closed room designed to restrict leakage. The 1.
minimum free volume shall be 2,000 cubic meters. exhausted from the reactor room and the Beam Port and
- 2. All air or other gas exhausted environment a Ventilation System shall be released to the environment Thermal Column Ventilation minimum of 26.5 meters above ground ground level.
TS 5.2 Reactor Coolant System TS Specifications
- 1. The reactor core shall be cooled by natural convective water flow.
demineralizer shall not extend more than 15 feet
- 2. The pool water inlet pipe to the demineralizer into the top of the reactor pool when fuel is in the core. The outlet pipe from the demineralizer shall be equipped demineralizer equipped with a check valve to prevent inadvertent inadvertent
- draining of the pool.
pumps shall be located no more than 15 feet
- 3. Diffuser and other auxiliary systems pumps below the top of the reactor pool.
- 4. All other piping and pneumatic tube systems entering the pool shall have siphon breakers and valves or blind flanges which will prevent draining more than 15 feet of water from the pool.
- 5. A pool level level alarm shall indicate loss of coolant if the pool level drops approximately one foot below normal level.
approximately
- Safety Analysis Report Rev. 2 UWNR Safe,ty 14-41 2008 Sept. 2008
5.3 Reactor Core and Fuel TS 5.3 Specifications Specifications 1.
- 1. TRIGA-FLIP TRIGA-FLIP Fuel The individual individual unirradiated FLIP fuel elemenfs elements shall have the following
- characteristics:
characteristics:
- a. Uranium Uranium content:
content: maximum maximum of 9 Wt-% enriched enriched to nominal 70% 70% Uranium 235. Hydrogen-to-zirconium atom ratio (in the ZrHJ:
- b. Hydrogen-to-zirconium ZrH,): nominal 1.6 H atoms to 1.0 Zr atoms.
- c. Natural Natural erbium content (homogeneously distributed):
distributed): nominal nominal 1.5 Wt-%.
- d. Cladding: 304 stainless steel, nominal 0.020 inch thick.
Identification: Top pieces
- e. Identification: pieces of FLIP fuel bundles will have characteristic characteristic markings to allow visual identification identification of FLIP fuel employed employed in mixed cores.
j
- 2. Standard TRIGA fuel
- The individual unirradiated unirradiated standard TRIGA fuel elements shall have the following characteristics:
a. a.. Uranium Uranium content: maximum maximum of 9.0 Wt-% of9.0 Wt-% enriched to a nominal 20% Uranium 235.
- b. Hydrogen-to-zirconium Hydrogen-to-zirconium atom ratio (in the ZrHx):
ZrHJ: nominal 1.7 H atoms to 1.01.0 Zr atoms.
- c. Cladding:
Cladding: 304 stainless steel, nominal 0.020 inch thick. UWNR Safety Analysis Report Report Rev. 2 14-42 14-42 Sept. 2008
- TS TS 5.4 Reactor Core
- Specifications Specifications
- 1. The core shall be an arrangement moderator of TRIGA uranium-zirconium arrangement ofTRIGA moderator bundles positioned uranium-zirconium hydride fuel-positioned in the reactor grid plate.
- 2. The TRIGA core assembly may be standard, FLIP, or a combination combination thereof thereof (mixed core), provided provided that any FLIP fuel be comprised of at least thirty-five (35)
(35) fuel elements, located in a contiguous, central region.
- 3. The reactor shall not be operated with a core lattice position vacant except for positions on the periphery of the core assembly.
positions
- 4. The reflector, excluding experiments and experimental excluding experiments experimental facilities, shall be water or a combination combination of graphite graphite and water.
TS 5.5 Control Elements TS Specifications
- 1. The safety blades shall be constructed of boral plate and shall have scram
- 3.
capability.. capability
- 2. The regulating regulating blade shall be constructed of stainless steel.
- 3. The transient transient rod shall contain contain borated borated graphite graphite or boron and its compounds compounds in a solid form as a poison in an aluminum aluminum or stainless steel clad. The transient control rod shall have scram capability capability and may incorporate incorporate an aluminum or air follower.
TS 5.6 Fissionable Material Material Storage Specifications Specifications
- 1. All fuel elements shall be stored in a geometrical geometrical array where the value of k-ofk-effective effective is less than 0.8 for all conditions of moderation.
- 2. Irradiated Irradiated fuel elements and fueled devices shall be stored in an array which will permit permit sufficient sufficient natural convection cooling by water or air such that the fuel exceed design values.
element or fueled device temperature will not exceed values .
- UWNR Safety Analysis Report Rev. 2 14-43 14-43 2008 Sept. 2008
- intentionally left blank.
This page is intentionally
- UWNR UWNR Safety-Analysis Safety.Analysis Report Report Rev. 2 14-44 14-44 Sept. 2008 2008
- TS 6.
TS ADMINISTRATIVE CONTROLS
- 6. ADMINISTRATIVE CONTROLS
- TS 6.1 TS 6.1 Organization Organization TS 6.1.1 TS 6.1.1 Structure The reactor facility shall be an integral part of the Engineering Physics Department of the College of Engineering Engineering of the University of Wisconsin-Madison. The reactor reactor shall be related related to the University structure as shown shown in Figure 14-1.
The Radiation Safety office performs audit functions for both the Radiation Radiation Safety Committee and the Reactor Safety Committee Committee and reports to both committees committees as well as to the Reactor Director. TS TS 6.1.2 6.1.2 Responsibility The Reactor Reactor Director Director is responsible for all activities at the facility, including licensing, security, emergency emergency preparedness, preparedness, and maintaining radiation radiation exposures as low as reasonably reasonably achievable. achievable. The reactor facility shall be under the direct control of a Reactor Supervisor Supervisor designated designated by the Reactor Reactor Director. The Reactor Supervisor Supervisor shall be
- responsible for assuring that all operations within the limits prescribed requirements requirements of the Radiation Committee..
Committee operations are conducted in a safe manner and prescribed by the facility license, procedures, and the Radiation Safety Safety Committee and the Reactor Reactor Safety
- UWNR UWNR Safety Analysis Report Rev. 2 14-45 2008 Sept. 2008
BOARD OF BOARD CHANCELLOR- MADISON CHANCELLOR-REGENTS OF REGENTS MADISON CAMPUS CAMPUS
- UNIVERSITY OF WISCONSIN UNIVERSITY WISCONSIN RADIATION SAFETY COMMITTEE RADIATION COMMITTEE University University Safety Department Department Radiation Safety Office Office CHAIRMAN ENGINEERING PHYSICS DEPARTMENT ENGINEERING DEPARTMENT (ANSI/ANS-15.4 Level 11))
(ANSI/ANS-1S.4
,REACTOR SAFETY COMMITIEE
,REACTOR COMMITTEE f---
- REACTOR DIRECTOR REACTOR DIRECTOR (ANSI/ANS-15.4 Level 2)
(ANSI/ANS-1S.4 SUPERVISOR (SRO) REACTOR SUPERVISOR REACTOR (ANSI/ANS-1 5.1 Level (ANSI/ANS-1S.1 Level 3) ALTERNATE SUPERVISORS SUPERVISORS (SRO) (ANSI/ANS-15.1 (ANSI/ANS-1S.1 Level Level 3) REACTOR OPERATORS REACTOR OPERATORS (RO)(RO)
*
(ANSI/ANS-15.1 (ANSI/ANS-1S.1 Level 4) Figure 14-1 Organization Organization Chart UWNR Safety Analysis Report Rev. 2 14-46 14-46 Sept. 2008
TS 6.1.3 TS 6.1.3 Staffing
- 1. The minimum
- 1. minimum staffing when the reactor reactor is not secured shall be:
- a. A licensed reactor operator operator in the control room (if senior operator senior operator licensed, may also be the person required in c).
- b. A second second designated designated person present at the facility capable of carrying carrying out prescribed written written instructions.
- c. A designated senior reactor reactor operator shall be readily available available at the facility or on call.
- 2. A list of reactor facility personnel by name and telephone number shall be readily available in the control room for use by the operator.
- 3. A licensed senior reactor operator operator shall be present at the facility for:
- a. Initial startup and approach approach to power.
- b. All fuel handling or control-element control-element relocations.
- c. Relocation of any in-core in-core experiment experiment with a reactivity worth greater
- than 0.7% ~KlK.
0.7% AK/K.
- d. Recovery from unplanned power reduction.
unplanned or unscheduled unscheduled shutdown shutdown or significant significant TS 6.1.4 TS Selection and Training Training of Personnel The selection, training, and requalification requalification of operations personnel shall meet or exceed exceed the requirements of ANSI/ANS-15.4-1988 ANSIIANS-15.4-1988 Sections 4-6. TS 6.2 TS 6.2 Review and Audit There shall be a Reactor Reactor Safety Committee which shall review and audit reactorreactor operations operations to assure that the facility is operated operated in a manner consistent consistent with public safety and within the conditions of the facility license license..
- UWNR UWNR Safety Analysis Report Rev. 2 14-47 2008 Sept. 2008
TS TS 6.2.1 6.2.1 Composition and Qualifications Qualifications
, The Committee shall be composed of a least six members, one of whom shall be a Health, Physicist from the University of Wisconsin Safety Radiation Safety Office. The Committee Radiation in the following disciplines:
Safety Department Committee shall collectively possess expertise expertise
*
- 1. Reactor
- 1. Reactor Physics;
- 2. Heat transfer transfer and fluid mechanics;
- 3. Metallurgy Metallurgy
- 4. Instruments A. Instruments and Control Systems;
- 5. Chemistry and Radio-chemistry; 5.
- 6. Radiation Radiation Safety.
TS 6.2.2 TS6.2.2 Charter and Rules
- The Committee Committee shall meet at least annually.
The Committee shall formulate written standards regarding the activities of of the full committee; committee; minutes, quorum, telephone polls for approvals not requiring a formal meeting, and subcommittees. TS TS 6.2.3 6.2.3 Review Function Function The responsibilities responsibilities of the Reactor Safety Safety Committee shall include, but are not limited to, the following:
- 1. Review and approval of experim~nts
- 1. experiments utilizing the reactor reactor facilities;
- 2. Review and approval of all proposed proposed changes changes to the facility, procedures, procedures, license, and technical specifications;
- 3. Determination Determination of whether a proposed change, test or experiment experiment would constitute unreviewed safety question or a change in Technical constitute an unreviewed Specifications; Specifications;
- 4. Review performance of plant equipment Review of abnormal performance equipment and operating UWNR UWNR Safety Safety Analysis anomalies Analysis Report Report Rev. 2 significance; and anomalies having safety significance; 14-48 14-48 2008 Sept. 2008
*
- 5. Review of unusual or reportable reportable occurrences occurrences and incidents which are
- reportable under 10 CFR Part 20 and 10 CFR Part 50. 50 .
- 6. Review of audit reports.
- 7. Review of violations of technical specifications, specifications, license, or procedures procedures and orders having safety significance.
significance. TS 6.2.4 TS Function Audit .Function A Health Physicist from the University of Wisconsin A Wisconsin Safety Department Department Radiation Safety Office shall represent the University Radiation Safety Committee and shall conduct an inspection inspection of the facility at least monthly to compliance with the regulations of 10 CFR Part 20. The services and assure compliance inspection inspection function of the Health Physics Office shall also be available to the Reactor Reactor Safety Committee, and will extend the scope of the audit to cover cover license, technical technical specification, specification, and procedure adherence. adherence. The committee shall audit operation operation and operational records of the facility. IfIf the committee committee chooses to use the staff of the Health Physics organization organization for the audit function, the reports of audit results will be distributed to the committee and included included as an agenda item for committee meetings meetings..
- Reactor staff shall perform annual reviews of the requalification program, the security plan, and the emergency plan and its implementing procedures.
TS 6.3 TS 6.3 Radiation Radiation Safety The Reactor Reactor Laboratory shall meet the requirements of the University requirements ofthe Radiation University Radiation Safety Regulations as submitted for the University Broad Broad License, License Number License, License Number 25-1323-01 and is subject 25-1323-01 subject to the authority of the state license. The Reactor Director shall have responsibility for maintaining radiation exposures as low as reasonably achievable achievable and for implementation implementation of laboratory laboratory procedure for insuring compliance regulations.. compliance with 10 CFR Part 20 regulations
- UWNR UWNR Safety Analysis Report Rev. 2 14-49 14-49 2008 Sept. 2008
TS 6.4 Procedures TS Written operating operating procedures procedures shall be adequate to assure the safety of operation of the ofthe
- reactor, but shall not preclude the use of independent judgement and action should the situation procedures shall be in effect for the following situation require such. Operating procedures items:
1.
- 1. Testing calibration of reactor operating instrumentation Testing and calibration instrumentation and controls, control rod drives, area radiation radiation monitors, and air particulate monitors;
- 2. Reactor Reactor startup, operation, and shutdown;
- 3. Emergency Emergency and abnormal abnormal conditions, including provisions for evacuation, evacuation, reentry, recovery, and medical support;
- 4. Fuel element element and experiment experiment loading or unloading;
- 5. Control rod removal or replacement; 5.
- 6. Routine maintenance maintenance of the control rod drives and reactor safety and interlock interlock systems or other routine maintenance maintenance that could have an effect on reactor safety;
*
- 7. Actions to be taken to correct malfunctions of correct specific and foreseen potential malfunctions of systems or components, including responses to alarms and abnormal reactivity changes; changes; and
- 8. Civil disturbances on or near the facility site.
Substantive Substantive changes changes to the above procedures procedures shall be made only with the approval of approval of the Reactor Safety Committee. Temporary changeschanges to the procedures procedures that do not change their original intent may be made by the Senior Operator Operator in control oror designated designated alternate. All such temporary changes shall be documented and subsequently reviewed by the Reactor Safety Safety Committee. UWNR Safety Analysis Report Rev. 2 14-50 14-50 Sept. 2008 2008
- TS 6.5 Experiment TS 6.5 Experiment Review Review and Approval
- 1. Routine experiments
- 1. experiments may be performed at the discretion responsible for operation operation without
- 2. Prior to performing proposed experiment discretion of the senior operator operator without the necessity of further review or approval.
performing any experiment experiment which is not a routine experiment, the experiment shall be evaluated by the senior operator operator responsible for operation. The senior senior operator shall consider the experiment experiment in terms of its effect on reactor operation and the possibility and consequences of its failure, including where significant, consideration consideration of chemical reactions, physical integrity, design life, proper cooling, interaction with core components, reactivity reactivity effects, and interactions with reactor instrumentation.
- 3. Modified routine experiments may be performed
- 3. performed at the discretion discretion of the senior operator responsible for operation operation without the necessity of further review or or approval provided that the evaluation accordance with Section evaluation performed in accordance 6.5(2) results in a determination determination that the hazards associated with the modified experiment are neither greater nor significantly routine experiment significantly different than those involved involved with the corresponding corresponding routine experiment which shall be referenced.
- 4. No special special experiment shall be performed until the proposed proposed experiment has beenbeen reviewed reviewed and approved by the Reactor Safety Committee.
- 5. Favorable evaluation of an experiment shall conclude
- 5. conclude that failure of the ofthe experiment will not lead directly to damage of reactor fuel or interference with movement of a control element.
- UWNR Safety Analysis Report Rev. 2 14-51-.
14-51-- 2008 Sept. 2008
I TS TS 6.6 6.6 Required Required Actions TS TS 6.6.1 6.6.1 Action to be Taken in Case of Safety Limit Violation In the event a safety limit is exceeded:
- W 1.
- 1. The reactor shall be shut down and reactor operation shall not be resumedresumed until authorized authorized by the NRC.
- 2. An immediate report of the occurrence occurrence shall be made to the Chairman, Chairman, Reactor Reactor Safety Committee, and reports shall be made to the NRC in accordance accordance with Section 6.7 of these specifications, specifications, and
- 3. A report shall be prepared prepared which shall include include an analysis of the causes and extent of possible resultant resultant damage, efficacy of corrective corrective action, and recommendations recommendations for measures measures to prevent prevent or reduce the probability of of recurrence.
recurrence. This report shall be submitted to the Reactor Safety Safety Committee Committee (RSC) for review and then submitted to the NRC NRC when authorization authorization is sought to resume operation operation of the reactor. TS 6.6.2 TS 6.6.2 Action to be Taken in the Event of an Occurrence of the Type Identified
- in 6.7.2(I)b.,
6.7.2(1)b., and 6.7.2(I)c. 6.7.2(1)c. In the event of an reportable occurrence (1.3.1) reportable occurrence (1.3.1) the following actions shall be taken:
- 1. The reactor shall be shut down.
1.
- 2. The Director or designated designated alternate alternate shall be notified and corrective action taken with respect to the operations operations involved,
- 3. The Director or designated designated alternate shall notify the Chairman of the Reactor Safety Committee,
- 4. A report shall be made to the Reactor Reactor Safety Committee which shall include an analysis of the cause of the occurrence, ofthe occurrence, efficacy of corrective recommendations for measures action, and recommendations measures to prevent prevent or reduce the probability probability of recurrence, and
- 5. A report shall be made to the NRC in accordance with Section
- 5. Section 6.7.2 of of these specifications.
specifications. 0 UWNR Safety Analysis Report Report Rev. 2 14-52 14-52 Sept. 2008
- TS 6.7 TS 6.7 Reports
- TS 6.7.1 TS 6.7.1 Operating Reports
- 1. An annual report covering the activities 1.
previous activities of the reactor facility during the calendar year shall be submitted (in writing to U.S. Nuclear previous calendar Regulatory Commission, Commission, Attn: Document Control Desk, Washington, DC DC 20555) within six months following the end of each calendar 20555) calendar year, providing the following information:
- a. A brief narrative experience (including narrative summary of (1) operating experience experiments experiments performed),
performed), (2) (2) changes changes in facility design, performance performance characteristics, characteristics, and operating procedures related to reactor safety and occurring occurring during the reporting period, and (3) results of surveillance tests and inspections;
- b. Tabulation of the energy output (in megawatt megawatt days) of the reactor, hours reactor was critical, and the cumulative cumulative total energy energy output since initial criticality;
- c. The number of emergency emergency shutdowns and inadvertent inadvertent scrams, including reasons therefor;
- d. Discussion of the major maintenance maintenance operations operations performed reactor and the reasons for any corrective maintenance
- e. A brief description, description, including performed during the period, including the effect, if any, on the safety of the operation operation of the maintenance required; summary of the safety evaluations including a summary evaluations of of changes in the facility or in the procedures and of tests and experiments experiments carried pursuant pursuant to Section 50.59 of 10 CFR Part 50;
- f. A summary summary of radiation exposures received received by facility personnel and visitors, including dates and time of significant significant exposures exposures and a summary of the results of radiation radiation and contamination contamination surveys performed performed within the facility; and
- g. A description description of any environmental environmental surveys performed performed outside the facility.
facility .
- UWNR Safety UWNR Safety Analysis Analysis Report Report Rev. 2 14-53.
14-53ý 2008 Sept. 2008
- h. A summary summary of the nature and amount of licensee as measured discharge; discharge;
~f radioactive effluents released or discharged to the environs beyond the effective measured at or prior to the point of such release or (summarized on a monthly basis)"
(1) Liquid Effluents (summarized basis)' released effective control of the
**0 discharged during the reporting period Liquid radioactivity discharged tabulated as follows:
(a) Total estimated radioactivity radioactivity released released (in curies) curies)... (b) The isotopic composition composition if greater 10'77 microcuries/cc greater than 1 x 10- microcuries/cc for fission and activation activation products. products. (c) Total radioactivity (in curies), released released by nuclide, during the reporting period based on representative representative isotopic analysis. (d) Average concentration at point of release (in microcuries/cc) Average concentration microcuries/cc) during the reporting reporting period and the fraction of the applicable applicable limit in 10 CFR Part 20. (e) Total volume (in gallons) of effluent water (including diluent) during periods of release. * (2) (summarized on a monthly basis) (2) Exhaust Effluents (summarized Radioactivity discharged Radioactivity discharged during the reporting period (in curies) for: (a) Gases. (b) Particulates Particulates with half lives greater greater than eight days. (c) (c) The estimated activity (in curies) discharged discharged during the reporting period, by nuclide, for all gases and particulates based on representative isotopic analysis and the fraction of the representative isotopic applicable applicable 10 CFR Part 20 limits for these values. (3) Solid Waste (3) (a) The total amount of solid waste packaged (in cubic feet). (b) The total activity activity involved involved (in curies). ((c) c) The dates of shipment shipment and disposition (if shipped off site).
- Auk UWNR Safety Analysis Report Rev. 2 14-54 14-54 2008 Sept. 2008
- 2. A report within 60 days after completion of startup testing of the reactor reactor
(in writing to the U.S. Nuclear Nuclear Regulatory Regulatory Commission, Attn: Document Control Desk, Washington, D.C. 20555) 20555) upon receipt of a new facility license or an amendment amendment to the license authorizing authorizing an increase in reactor reactor power level describing the measured measured values of the operating operating conditions conditions or characteristics of the reactor under the new conditions including: characteristics
- a. An evaluation of facility performance to date in comparison comparison with design predictions specifications, and predictions and specifications,
- b. A reassessment of the safety analysis submitted with the license application in light of measured operating characteristics characteristics when such measurements measurements indicate that there may be substantial substantial variance variance from from prior analysis.
TS 6.7.2 Special Reports
- 1. There shall be a report
- 1. report of any of the following not later than the following day by telephone or similar conveyance conveyance to the NRC Headquarters Headquarters Operation Operation Center, and followed by a written report describing the circumstances circumstances of the event and sent within 14 days to U.S. Nuclear Regulatory commission, Attn: Document Regulatory Document Control Desk, Washington,
- D.C. 20555:
- a. Any accidental release of radioactivity unrestricted radioactivity above permissible limits in unrestricted areas whether or not the release damage, personal injury, or exposure; release resulted in property
- b. Any violation of a safety limit; and
- c. Any reportable occurrences occurrences as defined in Section 1.3.1 of these specifications.
- 2. A written report within 30 days in writing to the U.S. Nuclear Regulatory Regulatory commission, Attn: Document Document Control Desk, Washington, D.C. 20555 of: of:
- a. Permanent Permanent changes in facility organization at Reactor Director or or Department Chair level.
- b. Any significant change in the transient transient or accident analysis as described in the Safety Analysis Report;
- UWNR UWNR Safety Analysis Analysis Report Report Rev. 2 14-55 14-55 2008 Sept. 2008
TS TS 6.8 6.8 Records TS TS 6.8.1 6.8.1 Records to be Retained for a Period of at least Five Years or for the Life of the Component 1. Component Involved if Less than Five Years
- 1. Normal reactor reactor facility operation (but not including supporting supporting documents
- such as checklists, log sheets, etc. which shall be maintained maintained for a period of at least one year),
- 2. Principal maintenance maintenance activities,
- 3. Reportable occurrences, Reportable occurrences,
- 4. Surveillance Surveillance activities activities required required by the Technical Specifications, Specifications,
- 5. Reactor
- 5. Reactor facility radiation and contamination contamination surveys where required by applicable regulations,
- 6. Experiments Experiments performed with the reactor,
- 7. Fuel inventories, inventories, receipts, and shipments,
*
- 8. Approved changes in operating procedures,
- 8. procedures,
- 9. Records of meeting and audit reports of the review and audit group.
TS TS 6.8.2 6.8.2 Records to be be Retained Retained for at Least One Cycle Operator qualification Operator re-qualification records. qualification and re-qualification TS TS 6.8.3 6.8.3 Records to be Retained Retained for the Lifetime of the Reactor Facility Annual reports which contain contain the information information in items 1I and 2 may be used as records for those items.
- 1. Gaseous and liquid radioactive
- 1. radioactive effluents released to the environs,
- 2. Offsite environmental environmental monitoring monitoring surveys required required by technical specifications, specifications,
- 3. Radiation Radiation exposures for all personnel personnel monitored,
*
- 4. Updated, corrected, and as-built drawings drawings of the facility.
UWNR Safety Analysis Report Rev. 2 14-56 14-56 2008 Sept. 2008
TS 77 REFERENCES TS REFERENCES
- 1. GA-9064, pages 3-1 to 3-23 1.
- 2. GA-9064, pages 3-1 to 3-23
- 3. NE Memo No.4,
- 3. No. 4, Report on Refueling the University of Wisconsin Nuclear Reactor, R. R. J.
1968, University of Wisconsin Department of Nuclear Engineering Cashwell, March 1968,
- 4. Core Test Program, UWNR Mixed TRIGA-FLIP TRIGA-FLIP Core (9 FLIP), R. R. J. Cashwell, July 1974, University of Wisconsin Department of Nuclear Engineering
- 5. Core Test Propgram, UWNR Mixed TRIGA-FLIP Core (15
- 5. (15 FLIP), R. J. Cashwell, February 1978, University of Wisconsin Department of Nuclear Engineering 1978,
- 6. Core Test Program, All FLIP Core, R. J. Cashwell, January 1980, University of Wisconsin Department of Nuclear Engineering
- 7. "Sagitta" refers to the bow of the element and means the maximum excursion of the clad surface from a chord connecting connecting the two ends of the clad surface.
- 8. Simnad and West, 1986
- 8. 1986
- 9. NUREG-1282
- UWNR UWNR Safety Safety Analysis Analysis Report Report Rev. 2 14-57 14-57 Sept. 2008 2008
- This page is intentionally intentionally left blank.
- UWNR UWNR Safety Safety Analysis Analysis Report Report Rev. 2 14-58 14-58 Sept. 2008 2008
*
- 15 15 15.1 15.1 FINANCIAL QUALIFICATIONS FINANCIAL Financial QUALIFICATIONS Financial Ability to Construct a Non-Power Not applicable applicable for renewal application.
Non-Power Reactor Reactor 15.2 15.2 Financial Ability to Operate a Non-Power Financial Non-Power Reactor Reactor The Reactor Laboratory is a part of the Engineering Physics Department of the College of of Engineering at the University ofWisconsin:-Madison. of Wisconsin-Madison. The teaching mission of the laboratory laboratory takes precedence precedence over the research research and service service missions; for this reason the primary primary fiscal support for operation of the facility is from the state-funded university budget. Reactor personnelpersonnel have instructional duties, such as teaching courses, setting up laboratory laboratory experiments, and assuring that equipment laboratories is operable. It thus becomes somewhat equipment used in the teaching laboratories difficult to allocate funding precisely precisely to just the operation operation of the reactor. In the following information no attempt is made to separate the instructional information instructional component of the budget from the reactor operations operations part of the budget. For instance, the Reactor Director is teaching two courses during the Spring 2000 semester, but his entire salary is included included in the budget below. Recently the operating budget of the Reactor Laboratory Laboratory has been considerably considerably higher than what would be predicted from the past history of the facility. This is due in large part to providing
- replacement replacement .of.of long-term long-term employees who are approaching approaching retirement age with younger younger workers in time to allow adequate training of the new employees. The total salary salary expenditures expenditures will be significantly after the retirements actually take place, but the estimate reduced significantly estimate below below uses the current funding level.
The total operating budget oftheof the laboratory laboratory is $388,000.
$388,000. Of this total, state instructional funding Ofthis covers $299,000, with the remaining remaining expenses split between grants ($48,000)
($48,000) and income generated by reactor services ($41,000). generated ($41,000). By expense expense category, category, the funds are spent for salary($253,000), salary($253,000), fringe benefits ($77,300), ($77,300), supplies supplies and expense expense ($24,000), ($24,000), and capital equipment equipment ($33,000). ($33,000). These numbers do not include the infrastructure Thyse infrastructure provided provided by the University University such as electrical janitor service, power, heating, janitor service, and health physics physics coverage. Further, the fringe benefits benefits included included above are not specifically specifically billed to the department in which the employee works for instructional funding, but come come from a campus-wide campus-wide fund, while the fringe benefits for salaries supported by non-instructional non-instructional funding are charged to the fund paying the salary. The instructional instructional funding is appropriated appropriated by the state. The administration administration of the university has been very supportive supportive of the reactor facility and continuation continuation of the Nuclear Nuclear Engineering Engineering curriculum. The fact that the application for renewal of the license is signed by the administration indicates this support. administration support.
- UWNR Safety Safety Analysis Analysis Report Rev. 0 15-1 15-1 April 2000
Much of the capital equipment equipment funding in recent years has come from the DOE program to
- update the instrumentation instrumentation and experimental experimental equipment on non-power reactors. In addition, a local utility company company has provided provided matching grants to support support instruction in traditional "fission "fission nuclear engineering", and this has contributed nuclear engineering", contributed to both the supply and expense and capital expense*and equipment budget. The combination combination of funding opportunities opportunities has resulted in the reactor and associated laboratories being in excellent condition. The reactor can continue to operate without the outside grant income, but it would not allow for upgrading upgrading the equipment equipment as has been done for the last 10 years. However, the present instrumentation and control systems are capable of present instrumentation of operate for another 20 years with no loss of function.
continuing to *operate Funds from services provided provided to persons persons outside the university have been steady for the last 10 10 increases in the last three years. It is expected years, with some increases expected that this funding will continue for at least the next five years. 15.3 15.3 Financial Financial Ability to De~ommission Decommission the Facility 1990 1; the University responded to 10 By letter dated July 19, 19901, 10 CFR Part 50.75 50.75 showing the expected decommissioning the reactor. At that time estimated expected cost of decommissioning estimated decommissioning decommissioning cost was $1,200,000 $1,200,000 in the year 2000. . Early in 1999 1999 the computations computations on which the funding plan was made were updated updated to extend the time of decommissioning decommissioning to 2020 and to incorporate incorporate more recent figures on the cost of disposal of radioactive debris from the decommissioning. decommissioning. The estimate is again based on placing the facility in condition for unrestricted unrestricted release three years after cessation of operations, operations, now now 0
- estimated as June 30, 2020. The result of this revised estimate estimate is a decommissioning decommissioning cost of of between between $3,000,000 and $8,000,000,
$8,000,000, with the wide range of cost based primarily upon the uncertainty uncertainty in disposal cost. Whereas the disposal cost was a minor part of the total decommissioning decommissioning effort costs for the original computation, it is now by far the'largest the' largest component component of the total cost. Nevertheless, Nevertheless, as a state agency, the funding plan remains to obtain the funding when necessary.
15.4 Reference 1.
- 1. Letter to USNRC Document Control Desk from R. J. Cashwell under Docket 50.156 50.156 19,1990, with attachment dated July 19,1990, attachment signed on behalf ofofth~
the Board of Regents. UWNR UWNR Safety Safety Analysis Report Report Rev. 0 15-2 15-2 April 2000
- e 16 16 OTHER LICENSE CONSIDERATIONS LICENSE CONSIDERATIONS 16.1 16.1 Prior Use of Reactor Components components in use at the University of Wisconsin Nuclear There are no components Reactor Laboratory that Nuclear Reactor have had prior use at any other facility or organization. conceivable that prior use organization. It is conceivable Reactor Laboratory systems at some future time.
components could be integrated into Reactor Appropriate analysis and reviews of component replacement Appropriate accordance to replacement will be conducted in accordance procedures and licensed technical applicable standards, regulations and facility procedures technical specifications. 16.2 16.2 Non-Power Reactors Medical Use of Non-Power Medical Nuclear Reactor Laboratory The University of Wisconsin Nuclear engaged nor licensed to conduct Laboratory is not engaged any activities for medical use of the facility. Future medical use of the Reactor Laboratory would license applications appropriate license be conducted pursuant to appropriate authorized by the applications and approvals as authorized Atomic Energy 1954 as amended. Energy Act of 1954 e* e Report Rev. 0 UWNR Safety Analysis Report UWNR 16-1 16-1 April 2000
*
( This page intentionally intentionally left blank.
- UWNR UWNR Safety Analysis Analysis Report Rev. 0 16-2 16-2 April 2000 April 2000
- 0
- Appendix Appendix A A.
A. Models A Calculation Calculation Methods Models Used for Calculations Atmospheric Release of Radioactivity Methods for Atmospheric Calculations in Sections 13.1.1.5 Sections 11.1.1.1.2 and 13.1.1.5 For Sutton's diffusion model, the maximum concentration (Xma) at any point downwind is given concentration (Xmax) as: (1) X (1) Xma- = 2Q 2Q (Reference I) 1) max er h-h 22 e1tj.L where ~ j is mean wind speed in meters/second meters/second QQ is release rate in Ci/second Ci/second hh is stack height in meters For the generalized generalized Gaussian Gaussian Plume Model, Model, the maximum maximum concentration concentration is given by the same (Reference 2, equation 8). equation (Reference 8). For calculations calculations in this report, the following values are used:
~ = lowest monthly average = = 3.54 meters/second meters/second hh == stack height above ground == 26.5 meters.
Using these numbers, equation (1) reduces to
(2) Xmax= Xmax = 9.42E-5 Q, where Xma Xmax is in I.Ci/ml j.LCi/ml Reference 2 presents Reference presents a method applicable applicable to release from buildings with zero stack height to approximate approximate release from leaks in a containment containment structure. structure. The relation given, as equation 4, is: (3) X= Q (7ra az+CA)iJ' where X, Q, Q, and [ij.L are defined defined as above, C is an empirical constant with a value between C between 0.5 to 2, and A A is the minimum building cross section. The a terms are concerned with atmospheric atmospheric dispersion, dispersion, which will be neglected neglected in this analysis, which will result in the equation; Q (4)X=~ (4)X=Q CAj.L
- UWNR Safety AnalysisAnalysis Report Rev. 2 A-1 A-I Sept. 2008
Inserting values for the UWNR facility used in the safety safety analysis analysis for FLIP fuel conversion, and
- using a value of 1 for C yields:
X=. (5) X-* . Q ,or
, or (1)(12,200ft (1 22 2/ft 2)(3.54m/sec)
)(12,200ft )(9.29E-2m 2ij(2)(3.54m/sec)
(6) X=2.49E-4 Q (6) Q with X of pLCi/ml X in units of\lCi/ml B. Sample Sample Calculations Supporting Supporting Section 11.1.1.1.2 11.1.1.1.2 The maximum release rate for Ar-41 activity is 13.3 pCi/second. \lCi/second. Using the ventilation ventilation system system of 9600 scfm, this activity is diluted to 2.94E-6 pCi/ml rated flow-rate of9600 \lCi/ml at the stack outlet. The concentration downwind, assuming the stack height, is calculated to be, from resulting maximum concentration equation (2) (2) (7) Xmax = (13.3E-6)(9.42E-5) (7) (13.3E-6)(9.42E-5) = 1.25E-9 \lCi/ml. ptCi/ml. If calculated using equation equation (6)(6) (which assumes zero stack height with building wake dilution), the resulting resulting value is (8) X- (13.3E-6)(2.49E-4) = 3.3 X= (13.3E-6)(2.49E-4) 1E-9 pCi/ml 3.31E-9 \lCi/ml It is obvious that the two methods used to calculate cannot both be applicable. calculate the above values cannot
- operated when the ventilation system is not in operation, the value in Since the reactor is not operated equation equation (7) (7) is more realistic, but the more conservative equation (8) is used in the text.
conservative value in equation C. C. Calculations Supporting Section 13.1.1.4 (1), Calculations Supporting (1), Whole Body Exposure concentration of the insoluble volatiles in the reactor The activity concentration determined by reactor room air was determined dividing released released activity by room volume. (9) ( ) A A _ 5.89E6 piCi = 2.95E-3 5.89E6 \lCi pCi/cm 3 2.95E-3 ~Ci/em3 9 V V 2.00E9 cm 33 2.00E9 em Since 3.7E4 dps = 1 IpCi, AIV= 109 y/sec-cm33
\lCi, A/V=
The maximum dose rate is calculated by assuming the room is equivalent to a hemisphere with a of 782 cm. In addition, the average gamma energy is 0.7 MeV, the attenuation radius of782 attenuation coefficient coefficient for air is 3.5E-5 cm',cm-\ and the flux-to-dose conversion y/cm 2/cm 2 -mr/min conversion factor is 4.2E4 y/cm2/cm2-mr/min Using the relationship relationship (10) DR (10) DR == 30S(1 exp( - R~>> 30S(1 -- exp(-RY,)) where where
- CY2
~
UWNR Safety UWNR Safety Analysis Report Report Rev. 2 A-2 A-2 Sept. 2008 2008
- DR = dose rate in mr/hr DR SS == Volumetric source strength in y/sec-cm22 R = outer R outer radius of hemisphere
~ = = attenuation attenuation coefficient for air, mr/hour.
yields a dose rate of 60 mrlhour. D. D. Calculations Supporting Section 13.1.1.4 (2), Dose to the Lungs The dose The dose toto the the lungs lungs was was calculated calculated by first assuming assuming uniform dispersal of the released volatiles in the laboratory volume, giving a concentration concentration of of (11) A _ 5.54E6 [Ci 2.75E-3 LCi/cm 33 . (11) A = 5.54E6 cm ~Ci3 = 2. 75E - 3 ~Ci/em 3 . V V 2.00E9 2.00E9 em 3 Since the "standard "standard man" breathes breathes 1.25 cubic meters of air per active hour, he would breathe approximately 0.21 cubic meters in 10 approximately 10 minutes, the assumed evacuation time. If this number is increased to 0.30 cubic meters to allow for excitement and stress, then his lungs would be increased exposed to an activity of of (12) A V (12) V =
= (.75E-3 ~C) pX) (3E5 (3E5 cmem 33)) = 825 ~Ci.
OCi. V V The dose to the lungs is then calculated calculated from the following expression to be 1 Rad Rad..
- (13) Dose (Rad) =ACR ~
(13) Dose (Rad) = - - L _ A A == Activity C m /=1 E F. E. (1 - e "/ where 1=1 Activity exposure (825 ~Ci) C = Conversion Conversion factors I_I (1 - e Aj 1. ItCi)
), where (3.7E4 P/sec-VQCi)(1.6E-6 (3.7E4 p/sec-~Ci)(1.6E-6 erg/MeV) erg/MeV) 1100erg/gm OOerg/gm -rad-rad R
R == lung retention factor (0.125 (0.125 is customary) customary) m m == mass of lungs (1000 oflungs (1000 grams) F. F j == fraction fraction of total activity Ei = energy E j = energy of of beta beta for for nuclide nuclide ii (MeV) (Me V)
'Xi = radioactive Aj = radioactive decay constant +
constant + biological release constant constant (6.7E-8 sec-) (6.7E-8 sec-I) t == time of exposure (assumed infinite) E. Calculations Supporting Section 13.1.1.5 Calculations Supporting Table 13.1 13.1.1.5 and Table 13.1 Release Release rate, rate, Q, Q, for an isotope isotope is the total (column E total quantity released to air (column E of of Table Table 13.1, 13.1, Chapter Chapter 13) 13) divided by the assumed release time. The The release release time used used in further calculations calculations is is the time for the time for the the ventilation ventilation system system (room air and beam port port and thermal thermal column exhaust exhaust systems) systems) to make make aa complete complete change change of air in the Reactor Reactor Laboratory. Laboratory .
- UWNR UWNR SafetySafety Analysis Analysis Report Report Rev. 2 A-3 A-3 Sept. 2008 2008
*
(14) TT, 2000 3 35.31 ft 3 60 sec = 1569 seconds (14) = 2000 m m 35.31 ft3 60 sec = 1569 seconds release scfin 2700 scfm 1I m 33 1 min Using the generalized generalized Gaussian Plume Model (equation (2)), (2)), and demonstrating demonstrating with data for Br-83, concentrations 83, concentrations released to unrestricted 13, Table 1 Column H) are calculated unrestricted areas (Chapter 13, calculated as shown below: ((15)
- 15) XX
)Br-83
= (0.0083Ci)(9.42E-5) r_83 =1569 1569 4.98E-1O p.Ci/ml (0.0083Ci)(9.42E-5) == 4.98E-10 1..1.
Cilml The remaining remaining isotope values are calculated calculated in the same manner. The activity release was also evaluated through use of equation (6). This calculation would be applicable to release of the activity through the building walls with the ventilation system not operating. Again using Br-83 as an example, and assuming the same release time as in the previous calculation calculation ( 16) XBr_83 (16) X (0.0083Ci)(2.49E-4) =- 1.32E-9
== (0.0083Ci)(2.49E-4) 1.32E-9 pCi/mlCilml Br-83 1569 1569 1..1.
greater than that evaluated by the Gaussian This value is a factor of 2.6 greater Gaussian Plume Model. All similar values in Chapter 13, 13, Table 1, Columns H and I may be multiplied by this factor for a more conservative conservative case. This calculation was done for the previous Safety Analysis Report using
- as the building dimensions dimensions only the minimum dimensions of the reactor laboratory, a room within the Mechanical Mechanical Engineering Engineering Building. This considerably underestimates underestimates the "wake effect" effect" of the actual building. The current analysis uses a value appropriate appropriate for the renovated Mechanical Mechanical Engineering Engineering Building (cross-sectional area of 12,200 Building (cross-sectional 12,200 ft2). ft2).
References References 1 Meteorology Meteorology and Atomic Energy, U. S. Dept of Commerce Weather Weather Bureau, Govt. Printing Office, Washington, DC July 1955 1955
- 3. F. A. Gifford, Jr, Atmospheric Atmospheric Dispersion Calculations Calculations Using the Generalized Generalized Gaussian Plume Model, Nuclear Safety, December December 1960
- 4. Calculation Calculation of Distance Factors for Power and Test Reactors, Reactors, (TID-14844), USAEC, March March 23, 1962 UWNR Safety Analysis UWNR Analysis Report Rev. 2 A-4 2008 Sept. 2008 '
*
- Appendix Appendix B B Supporting Supporting Documents Documents Task Task Order Order No. No.2 2 Under Under Master Master Task Task Agreement Agreement No. No. C96-175937 C96-17S937 Page Page 11 LMfTCO IMTCOfORM FORM PROC-ISI8b PROC-18I1b 07M 01/99 TASK TASKORDER ORDER NO. NO.2 2 UNDER UNDER MASTER MASTER TASK AGREEMENT NO.
TASK AGREEMENT NO. C96-175937 C96-175937 LOCKHEED LOCKHEED MARTIN MARTIN IDAHO IDAHO TECHNOLOGIES TECHNOLOGIES COMPANY COMPANY (LMITCO) (LMlTCO} 2525 2525 Fremont Fremont Avenue Avenue P. 0. P. O. Box 1625, Idaho 80x1625. Idaho Falls, Falls,IDID 83415-3521 83415-3521 OPERATING OPERATING UNDERUNDERU. U.S. S. GOVERNMENT COIITRACTNO. GOVERNMENT CONTRACT NO.DE-AC07-941D13223 DE*AC07*94lDU221 To: To: University of University ofWisconsin-Madison Wisconsin-Madison Effective EffectiveDate: Date: August26, August 26, 1999 1999 Research Research Administration Administration Completion Completion Date:Date: November November 1, I, 2001 2001 750 750 University UniversityAvenue Avenue Madison, Madison, WI WI 53706-1490 53706-1490 To: To: Tom Tom Handland Handland Ph: PI: R.I. R. J. Cashwell Cashwell This This Task Task Order OrderNo. No.2 2 isis awarded awarded tot to! 1.
- 1. Transfer Reactor Fuel Assistance Tl'llllsfer Reactor Fuel Assistance Subcontract SubcontractNo.No. C87-101251-002 C87-IOI1S 1-002 to to the the new new Master Master Task Task Agreement AgRCmentNo. No.C96-175937 C96-175937 as asTask OrderNo. 2.
TasIr;Order.No. 2. 2.
- 2. Extend Extend the theperiod period of performance to ofpcrformance November 1.
toNovember 2001.I. This 1,200 This extension extension isis retroactive retroactive to to November 1, November 1. 1998. 1998. 3.
- 3. Confim ConflIlJl the the Statement Statement of ofWork Work and and modification modification thereto thereto remain remain unchanged.
unchanged.
- 4.
- 4. Assignment: On Assignment: September30, On September 30, 1999, 1999, LMITCO's LMITCO'sprime primecontract with DOE contractwith DOE willwill expire.
expire. Thus, Thus, pursuant pursuant to the article tathe article in in the the General General Provisions Provisions entitled entitled"assignment""
"assignment"" thisthis Task TaskOrder Order isis assigned assigned toto Bechtel Bechtel BWXT 8WXTIdaho,Idaho. LLC,LLC, under under its its DOE DOE PrimePrime Contract ContractNo.No. DE-AC07-1DI3727, DE-AC07-IDI3727.
effective effective October October 1, 1. 1999. 1999. Procurement Proc:urement Agent Agent Lynda Lynda Keller Keller Telephone: (208) 526-5597 ( Cost: Cost: $0.00 SO.OO Ship Ship via: via: N/A N/A F.O.B/Trans.: /A Cash CashTerms: Terms: Net Net00 Days Days URfy* 1"!InsAddrm: Add.....: Signed: ý Lynda Keller Lynda Kdler Date LMITCO Lockheed Martin Idaho Teehnologles Company LMlTCO
. P.
P. 0. lc O. Box BoX 1625 1625 Idaho Idaho Falls, fall!, ID m83415-3521 Title: Procurement Agent 83415-J'21 T~C:~ Signed: Signed: ..,..-~..:2::!:~;:z...~I'::::l--=====~_ _...l(Q!.Lp-k~pJL..qqLL._ tohyI Subcoulkraet6s Subcontractor'sOftiat Otl\clal ' Date
'Di William Vance, Assistant WilliamJ.J. Vance, Assistant DeanDean Title:
Title: - RsmrDh Ruaatch &p I!I .*,*,q,*, . Sponsored ~l',\Im. - Return Return one Dno signed signed copy copy of ofthis chis Task TaskOrder Ordertoto Lynda LyndaKeller. Keller.
- UWNR UWNR Safety Safety Analysis Analysis Report Report Rev. Rev. 00 B-1 B-1 April 2000 April 2000
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- 0 UWNR UWNR Safety Analysis Report Rev. 0 B3-2 B-2 April 2000
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